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Multi-dimensional pool analysis of Phenix end-of-life natural circulation test with MARS-LMR code
Annals of Nuclear Energy 63 (2014) 309–316
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Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene
Multi-dimensional pool analysis of Phenix end-of-life natural circulation test with MARS-LMR code H.-Y. Jeong a,⇑, K.-S. Ha b, C.-W. Choi b a b
Sejong University, 98 Gunja-dong, Gwangjin-gu, Seoul 143-747, Republic of Korea Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejeon 305-353, Republic of Korea
a r t i c l e
i n f o
Article history: Received 23 May 2013 Received in revised form 31 July 2013 Accepted 2 August 2013 Available online 28 August 2013 Keywords: Sodium-cooled fast reactor Pool-type Multi-dimensional modeling Phenix end-of-life test Natural circulation MARS-LMR
a b s t r a c t The MARS-LMR code is a key system analysis tool for the development of a sodium-cooled fast reactor in Korea. The code has been successfully applied for the transient analysis of conceptual designs of SFR since 2007 mainly based on a one-dimensional approach. In recent studies, it was identified that one-dimensional modeling of a pool-type SFR has limitations on describing complicated thermal–hydraulic phenomena in pool regions at natural circulation conditions. In the present study, the natural circulation test performed in Phenix reactor by CEA has been analyzed with a multi-dimensional approach of MARS-LMR. Only the hot pool and the cold pool regions are modeled multi-dimensionally and other parts of the plant are described one-dimensionally in the analysis. Even though a very careful treatment of initial flow condition is required, this multi-dimensional modeling of pool regions results in quite accurate prediction of the temperature distributions measured at several points during the test when it is compared to the results with one-dimensional pool nodalization. It is suggested that a detailed modeling of pool regions is essential for the future analysis of pool-type SFRs. The multi-dimensional modeling capability can be enhanced through the improvement of the existing system code or by the combination of system code and CFD code. Ó 2013 Elsevier Ltd. All rights reserved.
1. Introduction At the dawn of the twenty-first century, the Generation IV International Forum (GIF) was founded for the development of Generation-IV (Gen-IV) nuclear energy systems which satisfy the goals of sustainability, economics, safety and reliability, proliferation resistance and physical protection (GIF, 2002). To search nuclear systems satisfying these Gen-IV goals the GIF evaluated more than 100 concepts and selected six most promising Gen-IV nuclear systems; Sodium-cooled Fast Reactors (SFR), Gas-cooled Fast Reactors (GFR), Lead-cooled Fast Reactors (LFR), Molten Salt Reactors (MSR), Very High Temperature Reactors (VHTR), and Super Critical Water-cooled Reactors (SCWR). Now, thirteen member countries are joining the GIF for the deployment of Gen-IV systems in a couple of decades. In Korea the superior safety characteristics of SFR and its possibility of increasing the sustainability of Uranium resources has boosted the research and development of SFR-related technology since 1997. As a result of these efforts the design concepts of KALIMER-150, KALIMER-600, and demonstration TRU burner have ⇑ Corresponding author. Address: Department of Nuclear Engineering, Sejong University, 98 Gunja-dong, Gwangjin-gu, Seoul 143-747, Republic of Korea. Tel./fax: +82 2 3408 4465. E-mail address: [email protected] (H.-Y. Jeong). 0306-4549/$ - see front matter Ó 2013 Elsevier Ltd. All rights reserved. http://dx.doi.org/10.1016/j.anucene.2013.08.010
been developed. Based on the previous experiences a conceptual design of prototype Gen-IV SFR is under development from 2012. It was recognized in the path of the design development that a flexible modeling tool is essential to evaluate several design options within the limited time schedule. To fulfill this request a new code named MARS-LMR was developed by modifying the existing MARS code (Jeong et al., 1999) which has been widely applied for the system analysis of a PWR in Korea. Actually, countries leading the development of SFR technologies have been using several SFR system analysis codes such as SAS4A/ SASSYS-1 (Fanning, 2012) in the United States, CATHARE-2 (Tenchine et al., 2012) in France, and etc. It is required to verify and validate the applicability and accuracy of the system analysis code for the use of design and licensing analysis. Recently, an IAEA coordinated research program (Tenchine et al., 2013) has been completed for the verification and validation (V&V) of the various system codes applicable to SFR analysis with the natural circulation (NC) data obtained in Phenix end-of-life (EOL) tests. Six one-dimensional codes and two three-dimensional codes have been benchmarked by eight organizations in this program. Korea joined the IAEA benchmark program on Phenix natural circulation test and evaluated the applicability of the MARS-LMR code to a pool-type SFR design (Jeong et al., 2011). In this study a one-dimensional modeling was applied for the prediction of thermal–hydraulic behaviors during a natural circulation condition.
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Through the one-dimensional analysis of Phenix natural circulation test it was found that one-dimensional modeling of a pool-type SFR has limitations on describing complicated thermal–hydraulic phenomena in pool regions at natural circulation conditions. Therefore, in recent study the Phenix end-of-life natural circulation test was analyzed with a multi-dimensional approach of MARS-LMR code and compared with one-dimensional analysis results (Jeong et al., 2012). In the present study, more detailed explanations on multi-dimensional analysis of Phenix natural circulation test with MARS-LMR for the hot pool and the cold pool regions are provided and the analysis results are delineated focusing on the effect of pool models. The simulation methodology described here is expected to enhance the accuracy in the evaluation of heat removal capability in future development of Gen-IV SFR systems. 2. Modeling of phenix NC test with MARS-LMR code In Korea efforts to develop a reliable and well-proven system analysis code for an SFR system have been continued since 1990s. One of the promising approaches was enhancing the modeling capability of the existing MARS code (Jeong et al., 1999) by reinforcing thermal–hydraulic models and reactivity feedback models dedicated to SFR systems, which resulted in the development of the MARS-LMR code (Ha et al., 2010). The MARS-LMR code is featured by the equation of state for sodium coolant properties, pressure drop correlations for wire-wrapped SFR core geometry, and heat transfer correlations related to a liquid metal. The code
has been applied successfully to the transient analysis of various conceptual designs such as KALIMER-600 and demonstration burner reactor. The MARS-LMR code is not required to be validated for all of the original models because it maintains the same governing equations and solution schemes adopted in the original MARS code. However, it is necessary to evaluate the applicability of the code to various conditions of SFR systems based on detailed assessment. In previous study (Ha et al., 2010) the loss of flow tests in EBR-II reactor were analyzed with the MARS-LMR and it was demonstrated that the code predicts the experimental data with great success. In the analyses the one-dimensional approach of MARS-LMR was applied, which resulted in quite good prediction of the transients in EBR-II, a small pool-type SFR. In other study (Tenchine et al., 2013) the predictability of MARS-LMR with one-dimensional representation of a rather large-sized pool-type SFR has been evaluated, which suggests there exist some limitations on describing the thermal–hydraulic behaviors during natural circulation. Therefore, the large hot pool and cold pool of Phenix are modeled with the MULTID component (KAERI, 2006) which is a special type of component similar to PIPE, VALVE, PUMP incorporated in the original version of MARS and also in MARS-LMR in the present study. 2.1. Phenix natural circulation test Phenix is one of the important prototypes in SFR developmental history. The reactor has been operated successfully for 35 years by the French Commissariat à l’énergie atomique (CEA) and the
Fig. 1. Schematic of Phenix reactor.
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‘Electricite de France (EdF) since 1973 to achieve the objectives of demonstration of fast reactor technology and of irradiation studies for innovative fuels and materials. The operation of Phenix was stopped in 2009 and CEA performed a set of end-of-life tests including thermal–hydraulic tests such as natural circulation. These EOL natural circulation tests provided a unique opportunity to evaluate the capability of SFR system analysis codes worldwide. As shown in Fig. 1, Phenix is a pool-type SFR of which primary system is composed of three primary pumps and six intermediate heat exchangers (IHXs) to transfer 563 MW of thermal power to steam generators. Just before the EOL natural circulation test the power level of Phenix was maintained at 350 MW with the operation of three primary pumps at 540 rpm to circulate the primary flow of 1840 kg/s. It should be noted that two secondary pumps in the intermediate heat transfer loops were working and, therefore, only four IHXs were active for heat transfer between the primary side and the secondary side. Two inactive IHX positions were replaced by closed cylindrical structures for cold pool temperature profile during the tests. It has been decided by the CEA to perform the natural circulation test at 120 MW power to guarantee the safety of the reactor. Therefore, the reactor power was decreased from 350 MW to 120 MW for about two and half hours to reach the initial test condition of power level. The speed of primary pumps was also
reduced from 540 rpm to 350 rpm stepwise to match the reactor power decrease. The main test was initiated by the manual decrease of feedwater flow rate followed by the dryout of steam generators (SGs). If the dryout of SGs (at reference time 0) was reached, the temperature difference between the primary and secondary sodium at the inlet of IHX became reduced. When this temperature difference reached 15 °C, the reactor was scrammed manually followed by the manual trip of primary pumps and the decrease of secondary pump speed to 110 rpm. After the stoppage of the primary pumps the decay heat from the core is removed only by the natural circulation flow rate in primary side. At about 3 h after the initiation of the test the opening of the SG casings causes a more active natural circulation and heat removal. 2.2. MARS-LMR modeling In the present study only the thermal–hydraulic aspect of the Phenix NC test is focused and the power change at the early stage of the test before the reactor scram due to reactivity feedback is not modeled with the kinetics models in MARS-LMR code. Therefore, the primary system including core, hot and cold pools, primary pumps, IHXs are modeled in detail, and the heat removal characteristics through the secondary and steam system are treated as boundary conditions. The important boundary conditions
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Fig. 2. Nodalization of Phenix reactor for MARS-LMR.
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Table 1 Simulation of Phenix at 120 MW condition with MARS-LMR. Plant parameter Core Inlet Temp., K Core Outlet Temp., K Core flow rate, kg/s 1st IHX In Temp., K 1st IHX Out Temp., K 2nd IHX In Temp., K 2nd IHX Out Temp., K 2nd flow rate, kg/s Hot pool level, m Cold pool level, m a b c
Target Operating Condition
Measured (averaged for first 5 s)
631.15 706.15 1284 705.15 633.15 581.15 705.15 760 10.03 9.72
633.19a 712.03b – 708.51 634.97c 580.45 703.65 – – –
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Reactor power, MW
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are core thermal power, primary pump speed, secondary flow rates, and secondary IHX inlet temperatures, which have been directly measured in the test. In Fig. 2, the nodalization scheme of the Phenix primary system used for MARS-LMR simulation is given. The core of 981 subassemblies is modeled into 7 flow groups of inner driver, outer driver, blanket driver, control rod, reflector, B4C shield, and hot assembly. Most parts of the system are described with one-dimensional approach except the hot and cold pools for which a multi-dimensional approach is adopted. The hot pool from the core outlet to the IHX inlets is nodalized into 8 nodes axially, 4 nodes radially, and 6 in azimuthal direction. In addition, the lower part of cold pool is discretized into 12 axial nodes, 1 radial node, and 9 azimuthal nodes considering the connection with IHXs and primary pumps. Basically, the numbers of radial and azimuthal nodes in pool regions are determined considering the location of onedimensional components such as IHXs and primary pumps, and the connection with the one-dimensional nodes. There are 4 radial nodes for the hot pool region. The bottom of inner-most hot-pool radial node 1 is connected with the outlet of inner core, outer core, blanket, and control subassemblies. The heat structures of core upper structure are located at the center of the radial node 1 from the second axial node to the top of the hot pool in axial direction. The next radial node 2 is connected with reflector subassemblies and node 3 with shield subassemblies. The top of the outer-most radial node 4 is connected to the inlet of IHXs. In the cold pool, a large core is located at the center region. Therefore, only 1 radial node is adopted to model the azimuthal motion of flow from IHXs to the pumps. As shown in the figure, a flow path for reactor vessel cooling is modeled quite complicatedly because this path governs the flow balance to core region and also is important for the temperature distribution in cold pool. Heat structures for fuel rods, IHX heat transfer tubes, metal structures in hot and cold pools, and reactor vessel are modeled in the simulation. For the analysis of Phenix natural circulation test with MARSLMR code the steady condition at 350 MW was simulated first. Then similar procedure for power reduction to 120 MW in real test was followed in the simulation by reducing the core power, primary pump speeds, and secondary flow rates to target values. In Table 1 the main system parameters obtained from MARS-LMR steady simulation are compared with the operating values which have been predicted with a design analysis tool by CEA. In the table some operating parameters which were measured directly in the Phenix EOL test are also provided. It should be noted that the calculated cannot be compared directly to the measured for some parameters because the locations of the measurement do not match exactly with the locations of simulation. Even though some
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At primary pump inlet. Averaged for several locations of subassembly outlet. Measured at IHX out elevation in cold pool (by thermocouple pole).
(b) Fig. 4. Phenix data of secondary flow rate and boundary condition for MARS-LMR.
parameters deviate from the target operating value or the measured one, all the major parameters for the initial state at 120 MW are predicted successfully well within the suggested uncertainties: ±5 K on temperatures, ±5% on flow rates. The main test of natural circulation is simulated by applying the imposed boundary conditions on core power, primary pump speed,
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secondary flow rate and the secondary temperature are used as boundary conditions in the calculation with MARS-LMR.
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3. Analysis results
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The analysis results of Phenix natural circulation test with MARS-LMR multi-dimensional modeling is described hereafter focusing on the directly measured parameters such as pump inlet temperature, subassembly outlet temperature, primary IHX inlet and outlet temperatures, and secondary IHX outlet temperature. The predictability of average core outlet temperature, which is not directly measured in the test, is also analyzed with related to the multi-dimensional flow behavior in the outlet plenum region. Some unique behaviors, which were not identified in one-dimensional analysis, are tried to be delineated in further detail. In Fig. 6, the short-term and the long-term behaviors of predicted pump inlet temperatures are provided. It should be noted that the relevant test data was measured from the thermocouple pole positioned in the cold pool, more specifically the thermocouple located at the level of the pump skirt suction. It was tried to get the core inlet temperature at the pump aspiration point in the derivation pipe inside the primary pump during the test. However, the temperature at this point is influenced by the ambient stratified hot sodium under natural circulation condition. The measured initial temperature of pump inlet is about 360 °C and it increases sharply after the dryout of SG to 406 °C followed by a slow decrease to a plateau temperature of 393 °C at 2000 s as shown in Fig. 6a. The calculated temperature for pump 1 inlet follows this trend of measured data quite accurately with the predicted peak temperature of 398 °C. It needs to be mentioned that the temperature distribution in the cold pool at this level has multi-dimensional behavior. Therefore, the calculated pump inlet temperatures for pump 1 and 3 do not predict the initial temperature peak at the very early phase of the transient. In Fig. 6b, it is shown that the long term trend of pump inlet temperature is successfully traced by the MARS-LMR simulation within the maximum deviation of 5 °C. The core outlet temperature is one of the most important parameters that should be estimated during natural circulation because it represents the effect of natural convection flow rate through the core region and it is also closely related to the safety aspect. In the test, core outlet temperatures were measured by thermocouples located at the upper plenum 10 cm above the outlet of selected subassemblies. In Fig. 7, the average core outlet temperature data is compared with the predicted temperatures at the uppermost node of subassemblies for three different zones of inner core, outer core, and blanket. The measured average core outlet temperature is a simple arithmetic mean value which does not take into account the influence of flow rate from each subassembly. Therefore, it does not represent the real evolution of temperature trend at a certain point of core outlet region. In the figure, it is found that the outer core temperature becomes higher than the inner core temperature for a short while between 700 s and 800 s. This trend of core outlet temperature is closely related to the flow rate through the subassemblies. After the reactor trip and pump trip, the flow rate through subassemblies decreases drastically before the natural circulation is established. At this stage, the flow through the inner core is maintained in upward positive direction. On the contrary, a reverse flow is formed in the outer core region for a while after 600 s in the simulation. When natural circulation recovers the positive flow through the outer core, the accumulated heat is transferred and the outlet temperature exceeds that of the inner core. It is not clear whether a reverse flow is formed or not in the real test because there is no test data on core flow rate for each region of core subassemblies and no measurement on outer core
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secondary inlet temperature, and secondary flow rate as mentioned previously. Figs. 3–5 are comparing the boundary conditions provided by CEA and the data given for the MARS-LMR simulation. As shown in Figs. 4 and 5 the interpolated values of
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Fig. 7. Profiles of the average core outlet temperature data and the predicted core outlet temperatures.
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temperature only. It is also found that the temperature decrease due to power decrease before reactor scram is over-estimated in the calculation. The reason of this over-estimated temperature reduction is found in the retarded increase of core inlet temperature before reactor scram as shown in Fig. 6a. It is estimated that the general trend of temperature increase due to the primary pump trip and the stabilization of temperature by the establishment of core-wide natural circulation flow is predicted successfully in the simulation. A more meaningful comparison of average core outlet temperature is found in Figs. 8 and 9. In Fig. 8, the predicted temperatures at three different azimuthal locations of upper plenum region just above the subassembly outlet are compared with the test data measured at 10 cm above the subassembly outlet. This location of upper plenum corresponds to the outlet of inner core, outer core, blanket, and control subassemblies. It is noted that the initial temperature at this elevation is predicted accurately in the simulation. It is also shown that the average core outlet temperature is reproduced accurately at upper plenum region for a certain period of time after a stable natural circulation is established. Fig. 9 compares the measured average core outlet temperature with the predicted upper plenum temperature averaged for upper plenum 1, 2 and 3. The maximum deviation of the predicted temperature from the measured one is about 10 °C when a natural circulation flow is stabilized as shown in the figure.
: Phenix data, core average : MARS-LMR, upper plenum average
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(b) Long-term Fig. 9. Comparison of the average core outlet data with the predicted upper plenum average temperature.
Fig. 10 is a more direct comparison of subassembly outlet temperatures measured and predicted for inner core subassembly zone. This figure can be considered as a typical example of real evolution of temperature history during the test. Even though the temperature decrease by a cold shock after reactor trip is over-estimated, the temperature peak after the primary pump trip is predicted quite accurately. The deviation between the measured and the predicted temperatures is within 5 °C during most of the transient. The MARS-LMR prediction of IHX inlet temperature in primary side is compared with the test data in Fig. 11. The measurement of IHX inlet temperature was obtained by the thermocouples located at the top of IHX inlet window. In Fig. 11, the temperature at the uppermost node of IHX tube bundle predicted by MARS-LMR shows an over-estimated initial decrease from 200 s to 460 s before reactor scram, which is again closely related to the initial decrease in core inlet temperature. After a stable natural circulation flow is established the long-term prediction of primary IHX inlet temperature follows the test data consistently as shown in Fig. 12. In the figure, the calculated temperatures for the hot pool nodes just before entering the IHX inlet windows are also provided. This result implies that the test data is between the predicted hot pool temperature and the IHX bundle temperature. It is conjectured that the discrepancy between the measured and the
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500 : Phenix data, average inner core : MARS-LMR, average inner core
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Fig. 12. IHX inlet temperature data compared with the predicted temperatures at the top of IHX tube bundle and at the neighboring hot pool node.
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Fig. 11. Prediction of short-term temperature at the top elevation of IHX bundle region.
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predicted is caused partly by the deficiency in the modeling of IHX inlet geometry including the shape of the inlet shroud. The short-term and the long-term behaviors of IHX primary outlet temperature are described in Fig. 13a and b, respectively. The relevant measurement was obtained by the thermocouple positioned in the cold pool region at the level of bottom IHX window. Therefore, the predicted temperatures for both the bottom of IHX tube bundle and the cold pool region near the IHX outlet window are depicted in Fig. 13. The initial increase of temperature due to loss of heat sink and the following sudden temperature drop by reactor scram, which was not found in one-dimensional calculation, is partially predicted in multi-dimensional analysis. The long-term data trend is well traced by the predicted temperature at the cold pool node neighboring the IHX outlet window. Especially, the cool down rate after the opening of the SG casing is predicted very accurately in Fig. 13b. In Fig. 14, the secondary IHX outlet temperature is predicted and compared with the test data. The initial temperature obtained from simulation is found to be over-estimated by 5 °C, which is presumed to be caused by the uncertainty in the estimation of heat transfer area and heat transfer rate through IHX tubes. The temperature decrease from 200 s to 460 s is also over-estimated in the simulation mainly due to the delayed increase in core inlet temperature as described earlier. When a natural circulation flow is
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(b) Long-term Fig. 13. Prediction of temperatures at the bottom of IHX tube bundle and cold pool node after outlet window.
established stably after 800 s, the measured data shows a continuous decrease up to the end of the test. This long-term behavior of IHX secondary outlet temperature including the enhanced cooling
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Through the analysis, it is shown that some multi-dimensional effects observed in the test, which was not predicted in the previous study with one-dimensional approach, can be reproduced with multi-dimensional function of MARS-LMR code. For example, the early peaks of pump inlet temperature and primary IHX outlet temperature are obtained in the present analysis, even though there are still some deviations from the test data. It is also shown that the short-term and long-term trend of core outlet temperature can be predicted quite accurately with the present approach. Further, it is evident that the general trend of main test parameters are described more correctly by treating large pool region multidimensionally and by distributing the heat structures more realistically. The active mixing predicted in the upper plenum region is mainly due to the flexible modeling of steel mass distribution in the pool regions. A more accurate modeling with increased number of multi-dimensional nodes and heat structures would enhance the prediction in future study. It is also expected that an improved prediction would be achieved if the decrease of temperatures at core outlet and IHX inlet before reactor scram could be removed in the simulation. It is suggested that multi-dimensional effects in pool-type reactor are investigated further for the validation of system code and its influence needs to be incorporated for the establishment of evaluation methodology for SFR safety analysis in the future.
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The authors wish to thank the CEA team who performed the Phenix EOL natural circulation test and provided the test data for the IAEA CRP. Special thanks to A. Vasile and P. Gauthe for arranging the international collaboration program and preparing plant data for analysis.
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(b) Long-term Fig. 14. Prediction of secondary IHX outlet temperature.
by the opening of SG casing at 10,500 s is correctly described within the maximum deviation of 10 °C by the MARS-LMR simulation. 4. Concluding remarks It is known that a one-dimensional system code has limitations on predicting complicated thermal–hydraulic phenomena associated with the natural circulation in a pool-type SFR. In the present study, the possibility of a more detailed description of natural circulation condition in a pool-type SFR is investigated by using the multi-dimensional function of MARS-LMR code for the evaluation of the Phenix EOL natural circulation test. In the analysis, only the hot pool and the cold pool are modeled with multi-dimensional nodalization, and the other part is treated with one-dimensional approach. The analysis is focused on thermal–hydraulic aspect of natural circulation test, thus, appropriate boundary conditions are imposed on the simulation.
GIF, 2002. A technology roadmap for Generation IV nuclear energy systems. In: GIF002-00, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum. Fanning, T.H. (Ed.), 2012. The SAS4A/SASSYS-1 Safety analysis code system. ANL/ NE-12/4, Nuclear Engineering Division, Argonne National Laboratory. Ha, K.S., Jeong, H.-Y., Cho, C.H., Kwon, Y.M., Lee, Y.B., Hahn, D., 2010. Simulation of the EBR-II loss-of-flow tests using the MARS code. Nucl. Technol. 169, 134–142. Jeong, H.-Y., Ha, K.S., Chang, W.P., Lee, K.L., 2011. MARS-LMR modeling for the posttest analysis of Phenix end-of-life natural circulation. Korean Nuclear Society Autumn Meeting, Geongju, Korea, October 27–28. Jeong, H.-Y., Ha, K.-S., Lee, K.-L., Chang, W.-P., Kim, Y.-I., 2012. Analysis of Phenix end-of-life natural convection test with the MARS-LMR code. ICAPP-2012, Chicago, USA, June 24–28. Jeong, J.J., Ha, K.S., Chung, B.D., Lee, W.J., 1999. Development of multi-dimensional thermal-hydraulic system code, MARS 1.3.1. Ann. Nucl. Energy 26, 1611–1642. KAERI, 2006. MARS code manual, Volume I: Code structure, System models, and Solution methods. KAERI/TR-2812/2004, Korea Atomic Energy Research Institute. Tenchine, D., Baviere, R., Bazin, P., Ducros, F., Geffraye, G., Kadri, D., Perdu, F., Pialla, D., Rameau, B., Tauveron, N., 2012. Status of CATHARE code for sodium cooled fast reactors. Nucl. Eng. Des. 245, 140–152. Tenchine, D., Pialla, D., Fanning, T.H., Thomas, J.W., Chellapandi, P., Shvetsov, Y., Mass, L., Jeong, H.-Y., Mikityuk, K., Chenu, A., Mochizuki, H., Monti, S., 2013. International benchmark on the natural convection test in Phenix reactor. Nucl. Eng. Des. 258, 189–198.
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Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene
Multi-dimensional pool analysis of Phenix end-of-life natural circulation test with MARS-LMR code H.-Y. Jeong a,⇑, K.-S. Ha b, C.-W. Choi b a b
Sejong University, 98 Gunja-dong, Gwangjin-gu, Seoul 143-747, Republic of Korea Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejeon 305-353, Republic of Korea
a r t i c l e
i n f o
Article history: Received 23 May 2013 Received in revised form 31 July 2013 Accepted 2 August 2013 Available online 28 August 2013 Keywords: Sodium-cooled fast reactor Pool-type Multi-dimensional modeling Phenix end-of-life test Natural circulation MARS-LMR
a b s t r a c t The MARS-LMR code is a key system analysis tool for the development of a sodium-cooled fast reactor in Korea. The code has been successfully applied for the transient analysis of conceptual designs of SFR since 2007 mainly based on a one-dimensional approach. In recent studies, it was identified that one-dimensional modeling of a pool-type SFR has limitations on describing complicated thermal–hydraulic phenomena in pool regions at natural circulation conditions. In the present study, the natural circulation test performed in Phenix reactor by CEA has been analyzed with a multi-dimensional approach of MARS-LMR. Only the hot pool and the cold pool regions are modeled multi-dimensionally and other parts of the plant are described one-dimensionally in the analysis. Even though a very careful treatment of initial flow condition is required, this multi-dimensional modeling of pool regions results in quite accurate prediction of the temperature distributions measured at several points during the test when it is compared to the results with one-dimensional pool nodalization. It is suggested that a detailed modeling of pool regions is essential for the future analysis of pool-type SFRs. The multi-dimensional modeling capability can be enhanced through the improvement of the existing system code or by the combination of system code and CFD code. Ó 2013 Elsevier Ltd. All rights reserved.
1. Introduction At the dawn of the twenty-first century, the Generation IV International Forum (GIF) was founded for the development of Generation-IV (Gen-IV) nuclear energy systems which satisfy the goals of sustainability, economics, safety and reliability, proliferation resistance and physical protection (GIF, 2002). To search nuclear systems satisfying these Gen-IV goals the GIF evaluated more than 100 concepts and selected six most promising Gen-IV nuclear systems; Sodium-cooled Fast Reactors (SFR), Gas-cooled Fast Reactors (GFR), Lead-cooled Fast Reactors (LFR), Molten Salt Reactors (MSR), Very High Temperature Reactors (VHTR), and Super Critical Water-cooled Reactors (SCWR). Now, thirteen member countries are joining the GIF for the deployment of Gen-IV systems in a couple of decades. In Korea the superior safety characteristics of SFR and its possibility of increasing the sustainability of Uranium resources has boosted the research and development of SFR-related technology since 1997. As a result of these efforts the design concepts of KALIMER-150, KALIMER-600, and demonstration TRU burner have ⇑ Corresponding author. Address: Department of Nuclear Engineering, Sejong University, 98 Gunja-dong, Gwangjin-gu, Seoul 143-747, Republic of Korea. Tel./fax: +82 2 3408 4465. E-mail address: [email protected] (H.-Y. Jeong). 0306-4549/$ - see front matter Ó 2013 Elsevier Ltd. All rights reserved. http://dx.doi.org/10.1016/j.anucene.2013.08.010
been developed. Based on the previous experiences a conceptual design of prototype Gen-IV SFR is under development from 2012. It was recognized in the path of the design development that a flexible modeling tool is essential to evaluate several design options within the limited time schedule. To fulfill this request a new code named MARS-LMR was developed by modifying the existing MARS code (Jeong et al., 1999) which has been widely applied for the system analysis of a PWR in Korea. Actually, countries leading the development of SFR technologies have been using several SFR system analysis codes such as SAS4A/ SASSYS-1 (Fanning, 2012) in the United States, CATHARE-2 (Tenchine et al., 2012) in France, and etc. It is required to verify and validate the applicability and accuracy of the system analysis code for the use of design and licensing analysis. Recently, an IAEA coordinated research program (Tenchine et al., 2013) has been completed for the verification and validation (V&V) of the various system codes applicable to SFR analysis with the natural circulation (NC) data obtained in Phenix end-of-life (EOL) tests. Six one-dimensional codes and two three-dimensional codes have been benchmarked by eight organizations in this program. Korea joined the IAEA benchmark program on Phenix natural circulation test and evaluated the applicability of the MARS-LMR code to a pool-type SFR design (Jeong et al., 2011). In this study a one-dimensional modeling was applied for the prediction of thermal–hydraulic behaviors during a natural circulation condition.
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Through the one-dimensional analysis of Phenix natural circulation test it was found that one-dimensional modeling of a pool-type SFR has limitations on describing complicated thermal–hydraulic phenomena in pool regions at natural circulation conditions. Therefore, in recent study the Phenix end-of-life natural circulation test was analyzed with a multi-dimensional approach of MARS-LMR code and compared with one-dimensional analysis results (Jeong et al., 2012). In the present study, more detailed explanations on multi-dimensional analysis of Phenix natural circulation test with MARS-LMR for the hot pool and the cold pool regions are provided and the analysis results are delineated focusing on the effect of pool models. The simulation methodology described here is expected to enhance the accuracy in the evaluation of heat removal capability in future development of Gen-IV SFR systems. 2. Modeling of phenix NC test with MARS-LMR code In Korea efforts to develop a reliable and well-proven system analysis code for an SFR system have been continued since 1990s. One of the promising approaches was enhancing the modeling capability of the existing MARS code (Jeong et al., 1999) by reinforcing thermal–hydraulic models and reactivity feedback models dedicated to SFR systems, which resulted in the development of the MARS-LMR code (Ha et al., 2010). The MARS-LMR code is featured by the equation of state for sodium coolant properties, pressure drop correlations for wire-wrapped SFR core geometry, and heat transfer correlations related to a liquid metal. The code
has been applied successfully to the transient analysis of various conceptual designs such as KALIMER-600 and demonstration burner reactor. The MARS-LMR code is not required to be validated for all of the original models because it maintains the same governing equations and solution schemes adopted in the original MARS code. However, it is necessary to evaluate the applicability of the code to various conditions of SFR systems based on detailed assessment. In previous study (Ha et al., 2010) the loss of flow tests in EBR-II reactor were analyzed with the MARS-LMR and it was demonstrated that the code predicts the experimental data with great success. In the analyses the one-dimensional approach of MARS-LMR was applied, which resulted in quite good prediction of the transients in EBR-II, a small pool-type SFR. In other study (Tenchine et al., 2013) the predictability of MARS-LMR with one-dimensional representation of a rather large-sized pool-type SFR has been evaluated, which suggests there exist some limitations on describing the thermal–hydraulic behaviors during natural circulation. Therefore, the large hot pool and cold pool of Phenix are modeled with the MULTID component (KAERI, 2006) which is a special type of component similar to PIPE, VALVE, PUMP incorporated in the original version of MARS and also in MARS-LMR in the present study. 2.1. Phenix natural circulation test Phenix is one of the important prototypes in SFR developmental history. The reactor has been operated successfully for 35 years by the French Commissariat à l’énergie atomique (CEA) and the
Fig. 1. Schematic of Phenix reactor.
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‘Electricite de France (EdF) since 1973 to achieve the objectives of demonstration of fast reactor technology and of irradiation studies for innovative fuels and materials. The operation of Phenix was stopped in 2009 and CEA performed a set of end-of-life tests including thermal–hydraulic tests such as natural circulation. These EOL natural circulation tests provided a unique opportunity to evaluate the capability of SFR system analysis codes worldwide. As shown in Fig. 1, Phenix is a pool-type SFR of which primary system is composed of three primary pumps and six intermediate heat exchangers (IHXs) to transfer 563 MW of thermal power to steam generators. Just before the EOL natural circulation test the power level of Phenix was maintained at 350 MW with the operation of three primary pumps at 540 rpm to circulate the primary flow of 1840 kg/s. It should be noted that two secondary pumps in the intermediate heat transfer loops were working and, therefore, only four IHXs were active for heat transfer between the primary side and the secondary side. Two inactive IHX positions were replaced by closed cylindrical structures for cold pool temperature profile during the tests. It has been decided by the CEA to perform the natural circulation test at 120 MW power to guarantee the safety of the reactor. Therefore, the reactor power was decreased from 350 MW to 120 MW for about two and half hours to reach the initial test condition of power level. The speed of primary pumps was also
reduced from 540 rpm to 350 rpm stepwise to match the reactor power decrease. The main test was initiated by the manual decrease of feedwater flow rate followed by the dryout of steam generators (SGs). If the dryout of SGs (at reference time 0) was reached, the temperature difference between the primary and secondary sodium at the inlet of IHX became reduced. When this temperature difference reached 15 °C, the reactor was scrammed manually followed by the manual trip of primary pumps and the decrease of secondary pump speed to 110 rpm. After the stoppage of the primary pumps the decay heat from the core is removed only by the natural circulation flow rate in primary side. At about 3 h after the initiation of the test the opening of the SG casings causes a more active natural circulation and heat removal. 2.2. MARS-LMR modeling In the present study only the thermal–hydraulic aspect of the Phenix NC test is focused and the power change at the early stage of the test before the reactor scram due to reactivity feedback is not modeled with the kinetics models in MARS-LMR code. Therefore, the primary system including core, hot and cold pools, primary pumps, IHXs are modeled in detail, and the heat removal characteristics through the secondary and steam system are treated as boundary conditions. The important boundary conditions
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Fig. 2. Nodalization of Phenix reactor for MARS-LMR.
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Table 1 Simulation of Phenix at 120 MW condition with MARS-LMR. Plant parameter Core Inlet Temp., K Core Outlet Temp., K Core flow rate, kg/s 1st IHX In Temp., K 1st IHX Out Temp., K 2nd IHX In Temp., K 2nd IHX Out Temp., K 2nd flow rate, kg/s Hot pool level, m Cold pool level, m a b c
Target Operating Condition
Measured (averaged for first 5 s)
631.15 706.15 1284 705.15 633.15 581.15 705.15 760 10.03 9.72
633.19a 712.03b – 708.51 634.97c 580.45 703.65 – – –
140
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Reactor power, MW
120 100 80 60 40 20 0 0
50
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Time, s Fig. 3. Power data of Phenix test and boundary condition for MARS-LMR.
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are core thermal power, primary pump speed, secondary flow rates, and secondary IHX inlet temperatures, which have been directly measured in the test. In Fig. 2, the nodalization scheme of the Phenix primary system used for MARS-LMR simulation is given. The core of 981 subassemblies is modeled into 7 flow groups of inner driver, outer driver, blanket driver, control rod, reflector, B4C shield, and hot assembly. Most parts of the system are described with one-dimensional approach except the hot and cold pools for which a multi-dimensional approach is adopted. The hot pool from the core outlet to the IHX inlets is nodalized into 8 nodes axially, 4 nodes radially, and 6 in azimuthal direction. In addition, the lower part of cold pool is discretized into 12 axial nodes, 1 radial node, and 9 azimuthal nodes considering the connection with IHXs and primary pumps. Basically, the numbers of radial and azimuthal nodes in pool regions are determined considering the location of onedimensional components such as IHXs and primary pumps, and the connection with the one-dimensional nodes. There are 4 radial nodes for the hot pool region. The bottom of inner-most hot-pool radial node 1 is connected with the outlet of inner core, outer core, blanket, and control subassemblies. The heat structures of core upper structure are located at the center of the radial node 1 from the second axial node to the top of the hot pool in axial direction. The next radial node 2 is connected with reflector subassemblies and node 3 with shield subassemblies. The top of the outer-most radial node 4 is connected to the inlet of IHXs. In the cold pool, a large core is located at the center region. Therefore, only 1 radial node is adopted to model the azimuthal motion of flow from IHXs to the pumps. As shown in the figure, a flow path for reactor vessel cooling is modeled quite complicatedly because this path governs the flow balance to core region and also is important for the temperature distribution in cold pool. Heat structures for fuel rods, IHX heat transfer tubes, metal structures in hot and cold pools, and reactor vessel are modeled in the simulation. For the analysis of Phenix natural circulation test with MARSLMR code the steady condition at 350 MW was simulated first. Then similar procedure for power reduction to 120 MW in real test was followed in the simulation by reducing the core power, primary pump speeds, and secondary flow rates to target values. In Table 1 the main system parameters obtained from MARS-LMR steady simulation are compared with the operating values which have been predicted with a design analysis tool by CEA. In the table some operating parameters which were measured directly in the Phenix EOL test are also provided. It should be noted that the calculated cannot be compared directly to the measured for some parameters because the locations of the measurement do not match exactly with the locations of simulation. Even though some
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At primary pump inlet. Averaged for several locations of subassembly outlet. Measured at IHX out elevation in cold pool (by thermocouple pole).
(b) Fig. 4. Phenix data of secondary flow rate and boundary condition for MARS-LMR.
parameters deviate from the target operating value or the measured one, all the major parameters for the initial state at 120 MW are predicted successfully well within the suggested uncertainties: ±5 K on temperatures, ±5% on flow rates. The main test of natural circulation is simulated by applying the imposed boundary conditions on core power, primary pump speed,
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secondary flow rate and the secondary temperature are used as boundary conditions in the calculation with MARS-LMR.
440 : Phenix, secondary IHX inlet temperature : MARS-LMR, secondary IHX inlet temperature
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3. Analysis results
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The analysis results of Phenix natural circulation test with MARS-LMR multi-dimensional modeling is described hereafter focusing on the directly measured parameters such as pump inlet temperature, subassembly outlet temperature, primary IHX inlet and outlet temperatures, and secondary IHX outlet temperature. The predictability of average core outlet temperature, which is not directly measured in the test, is also analyzed with related to the multi-dimensional flow behavior in the outlet plenum region. Some unique behaviors, which were not identified in one-dimensional analysis, are tried to be delineated in further detail. In Fig. 6, the short-term and the long-term behaviors of predicted pump inlet temperatures are provided. It should be noted that the relevant test data was measured from the thermocouple pole positioned in the cold pool, more specifically the thermocouple located at the level of the pump skirt suction. It was tried to get the core inlet temperature at the pump aspiration point in the derivation pipe inside the primary pump during the test. However, the temperature at this point is influenced by the ambient stratified hot sodium under natural circulation condition. The measured initial temperature of pump inlet is about 360 °C and it increases sharply after the dryout of SG to 406 °C followed by a slow decrease to a plateau temperature of 393 °C at 2000 s as shown in Fig. 6a. The calculated temperature for pump 1 inlet follows this trend of measured data quite accurately with the predicted peak temperature of 398 °C. It needs to be mentioned that the temperature distribution in the cold pool at this level has multi-dimensional behavior. Therefore, the calculated pump inlet temperatures for pump 1 and 3 do not predict the initial temperature peak at the very early phase of the transient. In Fig. 6b, it is shown that the long term trend of pump inlet temperature is successfully traced by the MARS-LMR simulation within the maximum deviation of 5 °C. The core outlet temperature is one of the most important parameters that should be estimated during natural circulation because it represents the effect of natural convection flow rate through the core region and it is also closely related to the safety aspect. In the test, core outlet temperatures were measured by thermocouples located at the upper plenum 10 cm above the outlet of selected subassemblies. In Fig. 7, the average core outlet temperature data is compared with the predicted temperatures at the uppermost node of subassemblies for three different zones of inner core, outer core, and blanket. The measured average core outlet temperature is a simple arithmetic mean value which does not take into account the influence of flow rate from each subassembly. Therefore, it does not represent the real evolution of temperature trend at a certain point of core outlet region. In the figure, it is found that the outer core temperature becomes higher than the inner core temperature for a short while between 700 s and 800 s. This trend of core outlet temperature is closely related to the flow rate through the subassemblies. After the reactor trip and pump trip, the flow rate through subassemblies decreases drastically before the natural circulation is established. At this stage, the flow through the inner core is maintained in upward positive direction. On the contrary, a reverse flow is formed in the outer core region for a while after 600 s in the simulation. When natural circulation recovers the positive flow through the outer core, the accumulated heat is transferred and the outlet temperature exceeds that of the inner core. It is not clear whether a reverse flow is formed or not in the real test because there is no test data on core flow rate for each region of core subassemblies and no measurement on outer core
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secondary inlet temperature, and secondary flow rate as mentioned previously. Figs. 3–5 are comparing the boundary conditions provided by CEA and the data given for the MARS-LMR simulation. As shown in Figs. 4 and 5 the interpolated values of
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Fig. 7. Profiles of the average core outlet temperature data and the predicted core outlet temperatures.
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temperature only. It is also found that the temperature decrease due to power decrease before reactor scram is over-estimated in the calculation. The reason of this over-estimated temperature reduction is found in the retarded increase of core inlet temperature before reactor scram as shown in Fig. 6a. It is estimated that the general trend of temperature increase due to the primary pump trip and the stabilization of temperature by the establishment of core-wide natural circulation flow is predicted successfully in the simulation. A more meaningful comparison of average core outlet temperature is found in Figs. 8 and 9. In Fig. 8, the predicted temperatures at three different azimuthal locations of upper plenum region just above the subassembly outlet are compared with the test data measured at 10 cm above the subassembly outlet. This location of upper plenum corresponds to the outlet of inner core, outer core, blanket, and control subassemblies. It is noted that the initial temperature at this elevation is predicted accurately in the simulation. It is also shown that the average core outlet temperature is reproduced accurately at upper plenum region for a certain period of time after a stable natural circulation is established. Fig. 9 compares the measured average core outlet temperature with the predicted upper plenum temperature averaged for upper plenum 1, 2 and 3. The maximum deviation of the predicted temperature from the measured one is about 10 °C when a natural circulation flow is stabilized as shown in the figure.
: Phenix data, core average : MARS-LMR, upper plenum average
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(b) Long-term Fig. 9. Comparison of the average core outlet data with the predicted upper plenum average temperature.
Fig. 10 is a more direct comparison of subassembly outlet temperatures measured and predicted for inner core subassembly zone. This figure can be considered as a typical example of real evolution of temperature history during the test. Even though the temperature decrease by a cold shock after reactor trip is over-estimated, the temperature peak after the primary pump trip is predicted quite accurately. The deviation between the measured and the predicted temperatures is within 5 °C during most of the transient. The MARS-LMR prediction of IHX inlet temperature in primary side is compared with the test data in Fig. 11. The measurement of IHX inlet temperature was obtained by the thermocouples located at the top of IHX inlet window. In Fig. 11, the temperature at the uppermost node of IHX tube bundle predicted by MARS-LMR shows an over-estimated initial decrease from 200 s to 460 s before reactor scram, which is again closely related to the initial decrease in core inlet temperature. After a stable natural circulation flow is established the long-term prediction of primary IHX inlet temperature follows the test data consistently as shown in Fig. 12. In the figure, the calculated temperatures for the hot pool nodes just before entering the IHX inlet windows are also provided. This result implies that the test data is between the predicted hot pool temperature and the IHX bundle temperature. It is conjectured that the discrepancy between the measured and the
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500 : Phenix data, average inner core : MARS-LMR, average inner core
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Fig. 12. IHX inlet temperature data compared with the predicted temperatures at the top of IHX tube bundle and at the neighboring hot pool node.
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Fig. 11. Prediction of short-term temperature at the top elevation of IHX bundle region.
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predicted is caused partly by the deficiency in the modeling of IHX inlet geometry including the shape of the inlet shroud. The short-term and the long-term behaviors of IHX primary outlet temperature are described in Fig. 13a and b, respectively. The relevant measurement was obtained by the thermocouple positioned in the cold pool region at the level of bottom IHX window. Therefore, the predicted temperatures for both the bottom of IHX tube bundle and the cold pool region near the IHX outlet window are depicted in Fig. 13. The initial increase of temperature due to loss of heat sink and the following sudden temperature drop by reactor scram, which was not found in one-dimensional calculation, is partially predicted in multi-dimensional analysis. The long-term data trend is well traced by the predicted temperature at the cold pool node neighboring the IHX outlet window. Especially, the cool down rate after the opening of the SG casing is predicted very accurately in Fig. 13b. In Fig. 14, the secondary IHX outlet temperature is predicted and compared with the test data. The initial temperature obtained from simulation is found to be over-estimated by 5 °C, which is presumed to be caused by the uncertainty in the estimation of heat transfer area and heat transfer rate through IHX tubes. The temperature decrease from 200 s to 460 s is also over-estimated in the simulation mainly due to the delayed increase in core inlet temperature as described earlier. When a natural circulation flow is
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(b) Long-term Fig. 13. Prediction of temperatures at the bottom of IHX tube bundle and cold pool node after outlet window.
established stably after 800 s, the measured data shows a continuous decrease up to the end of the test. This long-term behavior of IHX secondary outlet temperature including the enhanced cooling
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Through the analysis, it is shown that some multi-dimensional effects observed in the test, which was not predicted in the previous study with one-dimensional approach, can be reproduced with multi-dimensional function of MARS-LMR code. For example, the early peaks of pump inlet temperature and primary IHX outlet temperature are obtained in the present analysis, even though there are still some deviations from the test data. It is also shown that the short-term and long-term trend of core outlet temperature can be predicted quite accurately with the present approach. Further, it is evident that the general trend of main test parameters are described more correctly by treating large pool region multidimensionally and by distributing the heat structures more realistically. The active mixing predicted in the upper plenum region is mainly due to the flexible modeling of steel mass distribution in the pool regions. A more accurate modeling with increased number of multi-dimensional nodes and heat structures would enhance the prediction in future study. It is also expected that an improved prediction would be achieved if the decrease of temperatures at core outlet and IHX inlet before reactor scram could be removed in the simulation. It is suggested that multi-dimensional effects in pool-type reactor are investigated further for the validation of system code and its influence needs to be incorporated for the establishment of evaluation methodology for SFR safety analysis in the future.
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The authors wish to thank the CEA team who performed the Phenix EOL natural circulation test and provided the test data for the IAEA CRP. Special thanks to A. Vasile and P. Gauthe for arranging the international collaboration program and preparing plant data for analysis.
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References
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(b) Long-term Fig. 14. Prediction of secondary IHX outlet temperature.
by the opening of SG casing at 10,500 s is correctly described within the maximum deviation of 10 °C by the MARS-LMR simulation. 4. Concluding remarks It is known that a one-dimensional system code has limitations on predicting complicated thermal–hydraulic phenomena associated with the natural circulation in a pool-type SFR. In the present study, the possibility of a more detailed description of natural circulation condition in a pool-type SFR is investigated by using the multi-dimensional function of MARS-LMR code for the evaluation of the Phenix EOL natural circulation test. In the analysis, only the hot pool and the cold pool are modeled with multi-dimensional nodalization, and the other part is treated with one-dimensional approach. The analysis is focused on thermal–hydraulic aspect of natural circulation test, thus, appropriate boundary conditions are imposed on the simulation.
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