KEPCO International Nuclear Graduate School. ATLAS Facility and APR1400. A dissertation submitted in partial satisfaction of the

Transcription

1 KEPCO International Nuclear Graduate School A Comparative Study of DVI Line Break Accident between ATLAS Facility and APR1400 A dissertation submitted in partial satisfaction of the requirements for the deree Masters of Science in Nuclear Power Plant Enineerin by Erol Bicer 2015

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3 The dissertation of Erol Bicer is approved. Baik Se Jin Lee San Yon Jun Jae Cheon, Committee Chair KEPCO International Nuclear Graduate School 2015 ii

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5 TABLE OF CONTENTS TABLE OF CONTENTS... III LIST OF FIGURES... V LIST OF TABLES... VI ACKNOWLEDGEMENTS...VII VITA... VIII PUBLICATIONS... VIII ABSTRACT OF THE DISSERTATION... IX 1 INTRODUCTION DESCRIPTION OF ATLAS FACILITY MODELING OF MARS-KS ANALYSES MARS-KS GENERAL MODEL DESCRIPTIONS OVERVIEW OF APR1400 AND ATLAS MODELS CRITICAL FLOW MODELLING COUNTER-CURRENT FLOW LIMITATION DVI LINE BREAK ANALYSES RESULTS STEADY-STATE CALCULATIONS TRANSIENT CALCULATIONS CHRONOLOGIES OF THE ACCIDENT PRIMARY PRESSURES BREAK FLOW BEHAVIOURS LOOP SEAL CLEARING CORE COLLAPSED WATER LEVELS PEAK CLADDING TEMPERATURES CONCLUSION REFERENCES...31 iii

6 APPENDIX A CCFL MODEL...35 Appendix A.1 COFFICIENTS OF CCFL CORRELATION...35 Appendix A.2 CALCULATION PROCEDURE...36 APPENDIX B NODALIZATION OF MODELS...41 Appendix B.1 APR1400 NODALIZATION...41 Appendix B.2 ATLAS NODALIZATION...42 iv

7 LIST OF FIGURES Fiure 1-1 Safety Injection System of APR Fiure 2-1 Schematic Representation of ATLAS Facility... 6 Fiure 3-1 ECCS Components Layout Fiure 4-1 Primary Pressure Fiure 4-2 Break Flow Fiure 4-3. Loop Seal Clearin Fiure 4-4 Primary Pressure vs. Loop Seal Clearin Fiure 4-5 Core Collapsed Water Level Fiure 4-6 Axial Level Transmitters in ATLAS RPV Fiure 4-7 Peak Claddin Temperature Fiure 4-8 Peak Claddin Temperature vs. Break Size v

8 LIST OF TABLES Table 2-1 Major Scalin Parameters of ATLAS Facility... 7 Table 4-1 Steady-State Initialization Results Table 4-2 Accident in Chronoloical Order vi

9 ACKNOWLEDGEMENTS This work was supported by the 2015 Research Fund of the KINGS and performed with the data provided within the proram the 1st ATLAS Domestic Standard Problem (DSP-01), which was carried out by the Korea Atomic Enery Research Institute (KAERI) under the National Nuclear R&D Proram funded by the Ministry of Education, Science and Technoloy (MEST) of the Korean overnment. I am rateful to KINGS and the 1st ATLAS DSP-01 proram participants: KAERI for experimental data and the Council of the 1st DSP proram for providin the opportunity to publish the results. I would also like to deliver my sincere ratitude and appreciations to my committee chair Prof. Jun Jae Cheon, technical advisors Dr. Kim Taewan and Dr. Lee San Yon for their excellent uidance, motivation, immense knowlede and support throuhout my studies. I have been remarkably prosperous to have advisors who ave me the courae and inspiration to explore by myself as well as the direction when my steps come to a standstill. Besides my advisors, I would also like to thank my committee member Dr. Baik Se Jin for attendin to my defense, his comments, recommendations and editin this study. My sincere thanks also oes to my senior Tatu Alin for his reat contribution to this study. I also thank to my friends, KINGS staff and members for their best effort to make all students comfortable throuhout their life at KINGS. Finally, I would like to thank my parents, and my sisters for bein so patient and supportive throuhout my master deree study and my life in eneral. vii

10 VITA September 29, 1989 Born, Ardahan, Turkey 2013 B.S., Nuclear Enery Enineerin Hacettepe University Ankara, Turkey PUBLICATIONS Bicer, E., Tatu, A., Kim, T.W., and Ko, H.R., Scalin Analysis for DVI Line Break Accident of APR1400 based on ATLAS Experiment American Nuclear Society, NURETH-16, Chicao, IL, Auust 30-September 4, pp , (2015) viii

11 ABSTRACT OF THE DISSERTATION A Comparative Study of DVI Line Break Accident between ATLAS Facility and APR1400 by Erol Bicer Masters of Science in Nuclear Power Plant Enineerin KEPCO International Nuclear Graduate School, 2015 Professor Jun Jae Cheon, Chair The Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) facility has been desined and constructed by Korea Atomic Enery Research Institute (KAERI) to conduct interal effect tests for Advanced Power Reactor 1400 (APR1400). ATLAS has been scaled by usin the three-level scalin methodoloy suested by Ishii et al. [1] with the scalin ratio of 1/2 and 1/144 in lenth and flow area, respectively. Thus, the transients in ATLAS proress times faster than that the ones in the prototype, namely APR1400. In order to observe the scalability of ATLAS, Direct Vessel Injection (DVI) line uillotine break accidents for APR1400 and ATLAS has been analyzed by usin a system code, MARS-KS. The main idea of the analysis is to fiure out whether the phenomena at ATLAS, durin the accident, are oin to be reproduced at APR1400 or not. For a comparative study, primary ix

12 pressures, loop seal clearin, core collapsed water levels, and peak claddin temperatures have been investiated for both ATLAS and APR1400. Then the results were compared with ATLAS experiment data. In addition, a scalin relation for the coefficients of the Counter-Current Flow Limitation (CCFL) correlation has been calculated in order to compensate the scalin distortion induced by the different eometry at the fuel alinment plate. The final analyses results revealed that eneral plant behaviors, important phenomena and parameters durin the accident includin break flows, loop seal clearin and peak claddin temperatures are well reproduced in the analyses. It also indicates the scalability of ATLAS to APR1400 for the DVI line uillotine break accident. x

13 1 INTRODUCTION KAERI has constructed and operated the ATLAS which is a half-heiht, 1/144- flowarea scaled down interal test facility for APR1400 by usin Ishii and Kataoka s scalin method. Accordin to the law, the velocity at ATLAS is 2 times faster than one at APR1400. Thus, transient at ATLAS enerally proresses times faster than the prototype plant. The facility has the capacity to simulate a broad rane of Desin Basis Accidents (DBAs) [2]. Experiment results have been used for various purposes such as assessment of domestic codes, and developin emerency operatin strateies etc. DVI system is a special safety injection feature of emerency core coolin system in APR1400. While conventional safety injection systems are connected to the cold les, the DVI system is connected to the upper downcomer of the Reactor Pressure Vessel (RPV) as depicted in Fiure1-1 [3]. It is aimed to have no safety injection spillae durin cold le Loss of Coolant Accident (LOCA) by implementation of DVI system in APR1400 desin. It is also intended to reduce the pipin interconnections. In APR1400, four DVI lines are connected to the RPV. The safety injection flow of each DVI line comes from a Safety Injection Pump (SIP) and a Safety Injection Tank (SIT). SIP needs electric power to be operated. SIT model in APR1400 is similar to the accumulator in the typical Pressurized Water Reactors (PWR). In addition, there are two Emerency Diesel Generators (EDGs) to provide 1

14 electricity to the SIPs under loss of offsite power conditions and thus, each EDG is desined to supply electric power to two SIPs. Fiure 1-1 Safety Injection System of APR1400 Considerin the confiuration of the DVI system, it is clear that the worst sinle failure is the failure of an EDG since the two active safety injections are lost. Thus, the most sinificant accident reardin the DVI system is the uillotine break of a DVI line because the safety injections from both one SIP and one SIT could be lost. Therefore, from the safety injection point of view, the most limitin accident scenario for DVI line is a DVI line uillotine break with the failure of sinle EDG as sinle failure. In this case, the system has the least safety injection comin from three SITs and one SIP. In addition, importance of the accident was also indicated in Emerency Core Coolin Systems (ECCS) performance assessment under the provision of USNRC 10CFR50.46 [4]. 2

15 An interal effect test for the DVI line uillotine break accident has been performed by usin the ATLAS facility and the phenomena occurred durin the test was analyzed [5]. Benchmarkin MARS-KS over the experimental data was already performed under the framework of First Domestic Standard Problem for Code Assessment (DSP-01) for 100% DVI line break [6], DSP-02 for a 6-inch cold le break [7] and DSP-03 which is an onoin project for uillotine break of main steam line. The experimental results were also utilized as a reference data set for the DSP-01 [8]. More than 10 oranizations had participated in DSP-01. The important phenomena and eneral behavior of the system were well reproduced by the state-ofthe-art system codes such as RELAP5 [9], TRACE [10], and MARS-KS [11]. However, durin DSP-01 proram, investiations were mainly focused on the comparison between experimental data and simulation results of ATLAS facility. Thus, it is required to conduct a comparative study between ATLAS and APR1400 calculations in order to observe the scalability. This paper aims at presentin a comparative scalin analysis for the DVI line uillotine break of APR1400 based on the ATLAS experiment. It is also aimed to validate the scalability of ATLAS facility and the capability of MARS-KS code in simulatin the DVI line break accident. It is mainly focused on CCFL model which is sinificantly important for peak claddin temperature and core collapsed water level. ATLAS and APR1400 calculations were performed by usin MARS-KS system code and calculation results were compared with ATLAS experiment data which was obtained from DSP-01. 3

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17 2 DESCRIPTION OF ATLAS FACILITY The ATLAS facility is an interal effect test facility which is scaled-down from APR1400 with scalin ratios of 1/2 and 1/288 for lenth and volume, respectively. The facility has been desined to simulate thermal hydraulic states up to full pressure and temperature conditions of APR1400. As depicted in Fiure 2-1 [12], ATLAS is composed of two steam enerators (SGs), two hot les, four cold les and reactor coolant pumps, and a pressurizer which are as the same major component confiurations as APR1400. It incorporates all eometry characteristics of APR1400 and capable to simulate broad rane of scenarios such as Lare Break LOCA (LBLOCA), Small Break LOCA (SBLOCA), DVI Line Breaks, Main Steam Line Break (MSLB), Steam Generator Tube Rupture (SGTR) etc. The reactor pressure vessel includes the reactor core simulated with electrical heater rods, and interated annular downcomer. As mentioned before, the safety injection at APR1400 is established via the DVI system which is connected to the upper downcomer of the RPV. Therefore, four nozzles to model connection to the DVI line were installed at the upper downcomer. In addition, ATLAS has a connection to the safety injection system at each cold le which allows makin a comparative study for the DVI and cold le injection systems. It also incorporates the desin characteristics of Optimized Pressure Reactor 1000 (OPR1000). ATLAS also has break simulatin system, containment simulatin system and auxiliary systems. Break simulatin system allows 5

18 to simulate a variety rane of DBAs listed above and a quick openin valve, a break nozzle and various instruments were implement in the break simulatin system. Main purpose of the containment simulatin system is to collect the flow rate at breaks and maintain it in a specified back pressure. Fiure 2-1 Schematic Representation of ATLAS Facility Each SIT of APR1400 is equipped with a fluidic device in order to control the dischare flow rate passively. The fluidic device chanes the SIT operation mode from hih flow to low flow mode by usin its unique eometry [13]. Because of a 6

19 complicated eometry of the fluidic device, ATLAS models the function of the fluidic device usin a control valve. The scalin law that applied to ATLAS is the three-level scalin methodoloy which was developed by Ishii and Kataoka. Accordin to the scalin law, the halfheiht scalin will determine velocity 2 times faster at ATLAS than the one at APR1400. Thus, events at ATLAS enerally occur times faster than prototype plant. Table 2-1 Major Scalin Parameters of ATLAS Facility Parameters Scalin Law ATLAS Desin Lenth l 0R 1/2 Diameter d 0R 1/12 Area Volume 2 d or 1/144 2 l 0R d or 1/288 Core DT T 0R 1 Velocity Time Heat Flux 1/2 l 0R 1/2 l 0R 1/2 l 0R 1/ 2 1/ 2 2 Core Power l 1/2 2 0R d or 1/203.6 Pressure Drop l 0R 1/2 Flow Rate l 1/2 2 0R d or 1/203.6 The maximum core power of ATLAS is 1.96 MWth which represents 10 % of the scaled (1/203.6 scalin ratio) power. The core has electric heater rods and uide tubes with the same diameter and pitch to the reference desin. Volume was scaled by 1/288 to the prototype. The total number of fuel rods was scaled on the basis of an area ratio of 1/144. For heat transfer coefficient, same values were preserved by 7

20 keepin hydraulic diameter same as prototype. A list of scalin parameters is iven in Table

21 3 MODELING OF MARS-KS ANALYSES 3.1 MARS-KS MARS-KS (Multi-Dimensional Analysis of Reactor Safety) has been developed by KAERI by consolidatin the thermal hydraulic system code, RELAP5/MOD3.2, with interation of multi-dimensional subchannel analysis code, COBRA-TF [14]. MARS- KS has been written in FORTRAN90 which allows more convenient and effective development and maintenance of the code. A raphic user interface with real-time plot for minor edit variables is also one of the most important features of enhanced user-friendly environment. MARS-KS can be coupled with the three-dimensional reactor kinetics code, MASTER, and containment analysis codes such as CONTEMPT4 and CONTAIN. In addition to the capacity of both RELAP5 and COBRA-TF thermal hydraulic codes, MARS-KS has the capability to extend the analysis capacity by includin special thermal hydraulic models for tiht lattice core, CANDU, interal reactor, research reactor, and as-cooled reactor. One-dimensional features of MARS-KS have been used for the analyses, while MARS-KS has a capacity to analyze three-dimensional problems as well as typical one-dimensional cases. Analysis of both APR1400 and ATLAS were performed by usin MARS-KS code. 9

22 3.2 GENERAL MODEL DESCRIPTIONS OVERVIEW OF APR1400 AND ATLAS MODELS In order to analyze APR1400 reliably, it is very important to describe the eometrical characteristics of APR1400 appropriately. A base input deck was prepared for Lare Break LOCA (LBLOCA) Analysis of APR1400 desin by usin RELAP5/MOD3.3 code by Korea Institute of Nuclear Safety (KINS). Since it is actually impossible to collect all the data for APR1400, it is decided to collect eometrical information of APR1400 from this base input deck. The model for the scalin analysis has been developed with an idea where the experiment for DVI line uillotine break accident will be analyzed with APR1400 considerin the scalin parameters. Therefore, initial and boundary conditions at the experiment were applied to the APR1400 model after proper scalin In order to et the best results, several modifications were made in the steady state input deck. Time Dependent Volumes (TDVs) and Time Dependent Junctions (TDJs) were added to stabilize the primary pressure for steady-state initialization. All component scalin was performed in accordance with ATLAS model. Hot and averae channel were modelled to with proper heat structure models considerin the scalin laws. Variable and loical trips were set properly based on accident scenario. CCFL, model was applied to components such as core outlet, reactor vessel hot plenum nozzles, pressurizer sure line, lower plenum of steam enerators, reactor coolant pumps etc. Chockin and some other important models were applied to related components. Sinle junctions were replaced with valve components on the steam line. Reactor coolant pump data were revised and modifications were made. Power tables were adjusted properly for reactor core and reactor coolant pumps. Core 10

23 thermal power is 3988 MWt while reactor coolant pump net thermal power is 24.6 MWt. In restart input, proper loical and variable trips were added for reactor, reactor coolant pump, turbine, feed-water, safety injection etc. A TDV and a valve component were also added in order to simulate the containment buildin and break, respectively. Three new safety injection system models were added and connected to proper components in restart input. Safety injection tanks were modelled usin pipe components with proper scalin ratios. Decay heat curve was also revised for hot and averae channel. Nodalization of ARP1400 is depicted in Fiure B-1. ATLAS steady state and restart inputs were modified based on scalin laws. Modifications include those were made in APR1400 calculation in steady state and restart input. CCFL and chockin modellin were applied to the components where these models are sinificantly important. In order to et the best results for core collapsed water level and loop seal clearin, proper scalin corrections applied hot and averae channel in the reactor core. Safety injection system was aain modelled usin pipe components instead of accumulator models. Variable and loical trip cards were added after proper time scalin. Several new control inputs were also added in order to et the best results in the restart input of ATLAS calculation. Nodalization of ATLAS is depicted in Fiure B-2. The initial and boundary conditions as well as set points of reactor protection systems at the experiments were taken from the specifications for ATLAS DSP-01 [14]. Accordin to the specifications, the power iven to ATLAS was 8.0 % of the nominal power which equals to 1.56 MWt after subtractin initial heat loss of 85 kw in the experiment. Because the power at the experiment included the compensation 11

24 for heat loss durin the experiment, the power without the heat loss compensation was iven to the fuel rods in the APR1400 model. Fiure 3-1 ECCS Components Layout Since the failure of an EDG was assumed as sinle failure, the safety injection from both DVI lines next to the broken line was assumed to be lost. In addition, the broken DVI line also lost the injection from the SIT and SIP. Thus, three SITs and one SIP were modeled at the APR1400 and ATLAS models as depicted in Fiure 3-1. Other than ordinary modelin practicin with the accumulator model, both the SITs and sure lines were described by usin pipe components and a valve component has been used to simulate the fluidic device which allows the transition from hih to low flow modes. A control volume has been linked with a trip to reduce the flow area in SIT tank. When the water level ets lower than 3.27 meters, the trip is triered and 12

25 reduces the flow area in ATLAS calculation. Same method was also used in APR1400 calculation. SIT flow area is reduced when the water level is less than 6.25 meters. The initial pressure and temperature of the SIT were 4.2 MPa and 325 K, respectively, and the injection from the SIT starts at a pressure of 4.03 MPa. The SIP was modeled by usin time dependent volume and time dependent junction components with the mass flow rate table. The SIP injection is actuated by Low Pressurizer Pressure (LPP) trip at a pressure of MPa and a delay of 40.0 seconds was assumed in order to consider the delay due to the EDG start-up CRITICAL FLOW MODELLING A quick-openin valve has been employed to model the break at the DVI line. The break location is located at loop 2a which is depicted in Fiure 3-1. Containment buildins of APR1400 and ATLAS were modelled as TDVs. In eneral, containment simulatin systems are coupled with containment simulatin codes. Since this is a post-test calculation. It was decided to use TDVs as containment simulators. In order to et better results at the break flow rate, the critical flow was described by usin modified Henry-Fauske model. Since the APR1400 model was used to model the experiment at ATLAS, it is required to modify the dischare coefficient from a default value of 1.0 to a value suitable for the experiment. A previous study done by Ko and Kim [15] indicated that the proper value for the dischare coefficient is 0.75 based on sensitivity analyses done with MARS-KS calculations for the same ATLAS experiment. Since the scalin for the break mass flow rate was already done with the break flow area, it was decided to employ a dischare coefficient of 0.75 for both APR1400 and ATLAS analyses. 13

26 3.2.3 COUNTER-CURRENT FLOW LIMITATION The counter-current flow limitation (CCFL) can be present durin accidents at the upper core tie plate or fuel alinment plate, downcomer annulus, steam enerator tube support plates, and entrance to the tube sheet in the SG inlet plenum. In small-break LOCA analyses, it is very important to describe CCFL phenomena which occurs at the fuel alinment plate at the top of the reactor core. Durin the reflux condenser mode of SG, the condensate enerated at SG and flows back throuh the hot le to the fuel alinment plate. If the CCFL is not modeled appropriately, the condensate can directly fall into the reactor core and contribute to the core coolin. However, CCFL actually occurs there due to the small flow path throuh the fuel alinment plate and, as a result, the reactor core is prone has more chances to have a peak of the peak claddin temperature (PCT) due to less effective core coolin. Since CCFL phenomena are hihly correlated to the eometry, the coefficients for CCFL models should be selected considerin the eometry of the flow path. In MARS-KS, CCFL phenomena are modeled by usin the Bankoff correlation, which allows an interpolation between Wallis and Kutateladze forms [16]. Since the flow path at the fuel alinment plate has a small hydraulic diameter, the Wallis form of the CCFL correlation is more appropriate to be employed. In order to et the best results, a scalin relation has been developed between ATLAS and APR1400 models by usin the scalin law. Sensitivity analyses to determine two coefficients in the CCFL model has been carried out for ATLAS. Then the coefficients form ATLAS has been implemented to developed equation for APR1400 and, as a result, optimal coefficients are obtained as 0.4 and 1.0 for slope m and as intercept c, respectively. Calculation procedures to determine the coefficients can be found in Appendix A. 14

27 4 DVI LINE BREAK ANALYSES RESULTS 4.1 STEADY-STATE CALCULATIONS Steady state initialization results are summarized in Table 4-1 for experiment, APR1400, and ATLAS calculation. The column on the riht outermost side on the table represents the equivalent values. The values are obtained scalin down the APR1400 calculation results in order to observe the differences more conveniently. As indicated by Table 4-1, the steady state parameters predicted by APR1400 and ATLAS calculations make ood areements with correspondin parameters from ATLAS experiment. 4.2 TRANSIENT CALCULATIONS CHRONOLOGIES OF THE ACCIDENT Because of the scalin with the reduced heiht, the transient at ATLAS proress times faster than the one at APR1400. The accident was initiated at seconds at the experiment and ATLAS calculation which is equivalent to seconds at APR1400 thus, the DVI line uillotine break at APR1400 model was initiated at seconds in order to compare results with the experiment and calculation more conveniently. From this point, the reference time is oin to be the APR1400 time. 15

28 As mentioned before, the DVI line uillotine break occurred at seconds by openin a quick-openin valve at the upper downcomer. As soon as openin the break valve, the primary pressures bean to decrease and reached the set point of the Low Pressurizer Pressure (LPP) trip, MPa, at around seconds. Table 4-1 Steady-State Initialization Results Parameter Core Power [MW] Pressurizer Pressure [MPa] Core Inlet Temperature [K] Core Outlet Temperature [K] Cold Le Flow Rate [k/s] Secondary Pressure [MPa] Feedwater Temperature [K] Feedwater Flow Rate [k/s] SIT Pressure [MPa] SIT Temperature [K] SIT Level (%) Taret Value APR1400 (ATLAS) (MARS-KS) Primary System ATLAS (MARS-KS) Equivalent Value equivalent Secondary System 1.95 equivalent ECCS equivalent /94.9/ /94.9/ /94.9/ The scram sinal was enerated in 0.5 second from the LPP sinals. The main steam lines and secondary feed waters were isolated after the LPP sinal with delays of 0.1 second and 10.0 seconds, respectively. The core power was maintained 16

29 constant for around 5.7 seconds after the LPP sinal and started to follow the prorammed decay heat curve at seconds. At around seconds, the SIP was triered after a delay of 40.0 seconds after the LPP sinal. Three SITs started to deliver the ECC water at seconds when the upper downcomer pressure was 4.03 MPa in APR1400 calculation. In case of ATLAS calculation, ECC water started to be delivered at around seconds which is in the experiment. The calculation was terminated at 2,000 seconds. Table 4-2 Accident in Chronoloical Order Events ATLAS Experiment [time*, sec] APR1400 Calculation [time sec] ATLAS Calculation [time*, sec] Break Open Low Pressurizer Pressure Trip (LPP) Pressurizer Heater Trip Remarks P < MPa LPP Turbine Isolation LPP Reactor Scram & RCP trip Main Feedwater Isolation Safety Injection Pump Start LPP LPP LPP SIT Starts P < 4.03 MPa * Time of ATLAS is scaled to time of APR1400 by multiplyin a scalin ratio of The prediction of the chronoloy of the major events by MARS calculations were consistent with the experimental results. ATLAS calculation showed a very ood areement with experimental data. However, the SIT injection started earlier in APR1400 calculation due to relatively faster depressurization of the primary system. The chronoloy is summarized in Table

30 4.2.2 PRIMARY PRESSURES The time-traces of the primary pressures of experiment, APR1400 and ATLAS calculations are depicted in Fiure 4-1. As soon as openin the break valve, the primary pressure decreased rapidly due to the sudden loss of coolant inventory. Since APR1400 is a Liht Water Reactor (LWR) with U-tube steam enerators, typical scenarios of SBLOCA had happened such as blown-down phase. In this phase reactor coolant remained mostly liquid and rapid depressurization continued until the flashin of the coolant into steam was started in the natural circulation phase. After the depressurization rate was chaned, the primary pressure formed a plateau at a certain level between SIP injection and the loop seal clearin occurrence. Plateau formation started at 350 seconds when the circulation flow path of the two-phase mixture in the primary system was secured by the loop seal clearin. The plateau was ended at around 440 seconds and the primary pressure started decreasin rapidly aain. APR1400 and ATLAS calculations show ood areements up to the end of loop seal clearin as shown in Fiure 4-1. However, after the loop seal clearin the APR1400 calculation revealed very hih depressurization rate in pressure, comparin to the ATLAS experiment and calculation. There are two major reasons for the rapid depressurization of APR1400. First one is hih stored heat at ATLAS facility [12]. As a result of component scalin applied for ATLAS, it was indicated that the reactor pressure vessel had hiher stored heat than APR1400 and the correspondin scalin factor was 2.6. Second one is hiher thermal inertia number [12]. The thermal inertia number of ATLAS is 1.28 which was defined by a ratio of the thermal inertia of solid and liquid. Thus, it reveals that ATLAS has 30% hiher thermal inertia than APR

31 These analyses indicated that the depressurization of the primary system at ATLAS proresses slower than that at APR1400, and the calculation results revealed that the phenomena at the experimental facility are well predicted by the scalin analysis. Fiure 4-1 Primary Pressure BREAK FLOW BEHAVIOURS Fiure 4-2 compares the break flow rate between the experiment and MARS-KS calculations for APR1400 and ATLAS. Since there is a substantial pressure difference between the reactor pressure vessel and a tank simulates the containment, the critical flow condition was maintained throuhout the calculation time. The plot of the break flow shows a very clear transition of flow reime at the break from sinle-phase liquid, two-phase flow, and finally to sinle-phase vapor. The transition from two- 19

32 phase flow to sinle-phase vapor flow occurred by the depressurization followed by loop seal clearin. The default critical flow model of MARS-KS, modified Henry- Fauske model [17], was employed for the break flow modelin and Fiure 4-2 reveals that the characteristics of the break flow could be captured by the critical flow model with a dischare coefficient of Fiure 4-2 Break Flow The measured flow rate in the early stae at the experiment after the break is much hiher than calculation in the simulations. Thouh there is a lare difference between the experiment data and calculations, trend of the break flow rates show ood areement especially after the loop seal clearin. It is always hard to measure the flow rate in experimental facilities, especially when it is two phase flow. In case of ATLAS, containment simulatin system was 20

33 desined in a way that it has the capability to measure flow rates in separated vessels for two-phase breaks with two different measurement techniques [18]. The idea is that water and vapor are completely separated in a separatin vessel. After the separation of phases two different flow rate measurin systems apply to different phase. For water, load cell flowmeter is used and it was found out that in hiher flow rate flows cannot be measured reasonably because of delay time drained water. Thus, the break flow rate data in the early stae of the experiment has a lot of uncertainties. Once the correct measurement technique has applied, better results can be achieved for break flow rate LOOP SEAL CLEARING Loop seal clearin is the cleanin out of the water accumulated at crossover les durin the cold-le loss of coolant accidents by steam which has a sinificant influence on core durin reflood phase. The variation of the collapsed water level in the vertical intermediate les is presented in Fiure 4-3. The condensate was enerated at the SG durin the reflux condenser mode and accumulated in the intermediate les. Due to the accumulation of the condensate, the natural circulation flow path between the upper plenum and cold le nozzle at the downcomer could not be established successfully. Thus, the pressure difference between the upper plenum and downcomer increased. When the pressure difference became lare enouh to push the accumulated condensate out at the intermediate le, the condensate was removed at once and then a stable natural circulation path was established. Fiure 4-3 indicates that two loop seals of one loop were cleared at around 290 seconds which is also followed by a simultaneous clearin of remainin two loop 21

34 seals at the other loop at around 310 seconds in the experiment. This delay is caused by different depressurization rates and non-uniformly distributed water level in annulus reion of downcomer. It was also found out that the APR1400 and ATLAS calculations could capture the loop seal clearin phenomena at the same time as in the experiment. As you can see in the Fiure 4-3, water levels of intermediate les are different at experiment. This is due to the different locations of level sensors. However, most important thin here to point out is the fact that loops seal clearin time showed a ood areement between experimental data and calculations of ATLAS and APR1400. Fiure 4-3. Loop Seal Clearin In addition, the primary pressure decreased when the loop seal clearin occurred because of no more pressure build-up in the system. In order to see the effect 22

35 of loop seal clearin evidently, primary pressure and cross over le levels were sketched in the same raph. Fiure 4-4 shows the formation of the plateau reion clearly and reveals that the depressurization of the primary system started aain riht after the loop seal clearin. Fiure 4-4 Primary Pressure vs. Loop Seal Clearin CORE COLLAPSED WATER LEVELS Core collapsed water level has sinificant influence upon peak claddin temperature. A comparison of core collapsed water level between experimental data and predicted calculations of APR1400 and ATLAS is depicted in Fiure 4-5. Referrin to the fiure, two phase condition of coolant was observed after the break occurrence until the end of the accident. As it can be seen in the fiure, core collapsed water level exceeds the initial after the break occurred. This is caused by a sensor error at the 23

36 experiment. The core level started to decrease riht after the break occurred and core started to uncover which caused the peak claddin temperature to increase. At around 290 seconds loop seal cleared at one of the loops and made the collapsed water level increase aain which is also followed by the clearance of the other loop at around 310 seconds. Fiure 4-5 Core Collapsed Water Level Once the loop seal clearin is finished core collapsed water level started to decrease aain and, as it can be seen from Fiure 4-5, collapsed water level is lower in APR1400 calculation. This is because of faster depressurization which is described in section Then primary pressure continued to lower and has become lower than 4.02 MPa which caused the SITs to inject emerency core coolin water into the core. Faster depressurization of APR1400 also caused the SITs to start injectin emerency 24

37 core coolin water earlier than ATLAS experiment and calculation. After SIT injection the collapsed water level of both experiment and calculations showed ood areement. Injectin emerency core coolin water to by SITs made core collapsed water level increase aain and core started to recover. Fiure 4-6 Axial Level Transmitters in ATLAS RPV One important thin should be explained at this point about Fiure 4-5. Accordin to SB-DVI-08 experiment data, as it can also be seen in the fiure above, the initial core collapsed water level of ATLAS is 3.01 m. However, KAERI/TR- 4316/2011 documentation indicates that the related level transmitter (LT-RPV-01) position of active core level is located at 2.91 m. It is considered as a sensor error 25

38 durin the experiment [19]. Axial level transmitters in Reactor Pressure Vessel (RPV) is depicted in Fiure 4-6. Considerin the scalin analysis between ATLAS and APR1400, it was decided to shift up the APR1400 results in order to see the results conveniently PEAK CLADDING TEMPERATURES Durin the pressure build-up at the upper plenum, the water levels at downcomer and core were totally different from each other due to the different pressure at downcomer and upper plenum. Since the pressure at the upper plenum was hiher than that at the downcomer, the core had a lower water level than the downcomer. Such an inequality in the water levels ot bier because of the continuous pressure build-up at the upper plenum. As a result, the core started to be uncovered and the claddin temperature of the uncovered fuel rods increased. Peak Claddin Temperature (PCT) of experiment and calculations are shown in Fiure 4-7. At the experiment PCT reached to K at around 290 seconds in the outer reion of heated rods while K and K temperatures were observed as PCTs in ATLAS and APR1400 calculations, respectively. The differences amon the heater rods at ATLAS experiments are also not sinificantly different. As shown in Fiure 4-7, PCTs of ATLAS and APR1400 calculation were observed almost at the same time with experiment. In this stae, the loop seal clearin plays very important role in coolin down the uncovered fuel rods by floodin the core. The heat-up of the uncovered fuel rods continued until the core level was brouht above the top of the active fuel as a result of the loop seal clearin. 26

39 Fiure 4-7 Peak Claddin Temperature When the loop seal clearin occurred, the pressure difference between the downcomer and the upper plenum decreased substantially by establishin the natural circulation path between them. Because of the equilibrium in pressure, the water level at the core increased, whereas the water level in the downcomer decreased as shown in Fiure 4-5. The water level increase at the core helped floodin the uncovered fuel rods and prevented the claddin temperature to increase further. As a result, the claddin temperature made a peak and decreased aain, as depicted in Fiure 4-7. As it can be seen in the fiure the PCT decreased relatively faster in the APR1400 calculation. This is because of the scalin distortion in thermal inertia discussed in section In eneral, the phenomena at ATLAS experiment related to the loop seal behavior were reproduced appropriately in the MARS-KS calculations for 27

40 APR1400 and ATLAS. Accordin to APR1400 Desin Control Document (DCD), maximum PCT is K [20] for DVI line break accident which is hiher than the hihest PCT was observed in the calculations and present data. Hiher PCT in DCD implies that more conservatism has been included in the methodoloy. In addition, a raph of PCT versus break size is depicted in Fiure 4-8 [21] which reveals that hihest claddin temperature is achieved on DVI line amon other SBLOCA analyses. Fiure 4-8 Peak Claddin Temperature vs. Break Size 28

41 5 CONCLUSION A comparative study for a DVI line uillotine break accident has been carried out in order to address similarity between APR1400 and its scaled-down experimental facility, ATLAS. Analyses were conducted by simulatin an interal effect test for the DVI line break at ATLAS usin the MARS-KS calculation with the APR1400 and ATLAS models. Primary pressures, break flow rates, loop seal clearins, core collapsed water levels, and peak claddin temperatures were comparatively investiated in order to have a better understandin in DVI line break accident scenario. It is known that the behavior of primary pressure is hihly dependent on the dischare coefficients at the break. Thus, a modified Henry-Fauske model was used in both calculations. Thus the related phenomena showed ood areements between experiment results and calculations. In addition, a scalin relation calculation was carried out between APR1400 and ATLAS in order to find the coefficients of countercurrent limitation model since the prediction of peak claddin temperatures is hihly dependent on it. PCTs predicted accurately by usin the calculated coefficients by MARS-KS code. After comparin the calculated hihest PCT with the one that presented in DCD of APR1400, it is found out that more conservatism has been included than this study and present data. The results also indicated that the eneral thermal hydraulic behavior at ATLAS durin the experiment was well reproduced by 29

42 both APR1400 and ATLAS calculations. Especially, the loop seal clearin, one of the most important phenomena in determinin the PCT, was well predicted by both APR1400 and ATLAS calculations. It is also shown that MARS-KS code is capable to simulate DVI line break accident reasonably. Althouh the ATLAS experiment and calculation indicated slower depressurization than APR1400, eneral behaviors showed ood areements. The slower depressurization had been predicted from hiher thermal inertia of ATLAS and such a scalin distortion could be analyzed by usin scalin law that applied for the experimental facility. 30

43 6 REFERENCES [1] M. Ishii and I. Kataoka, Similarity Analysis and Scalin Criteria for LWRs under Sinle Phase and Two-Phase Natural Circulation, NUREG/CR-3267, ANL 83-32, Aronne National Laboratory, (1983). [2] K.Y. Choi, et al., Simulation Capability of the ATLAS Facility for Major Desin- Basis Accidents, Nuclear Technoloy, 156 (3), pp , (2006). [3] KEPCO, Status Report 83 - Advanced Power Reactor 1400 MWe (APR1400), pp. 12, (2011). [4] USNRC, Acceptance Criteria for Emerency Core Coolin Systems for Liht- Water Nuclear Power Reactors, 10CFR50.46, (2014). [5] K.Y. Choi, et al., Experimental Simulation of a Direct Vessel Injection Line Break of the APR1400 with the ATLAS, Nuclear Enineerin and Technoloy, 41 (5), pp (2009). [6] Y.-S. Kim and et. al., "First ATLAS Domestic Standard Problem (DSP-01) For the Code Assessment," Korean Nuclear Society, (2010). [7] Y.-S. Kim and e. al., "Second ATLAS Domestic Standard Problem (DSP-02) For a Code Assessment," Korean Nuclear Society, (2013). [8] Y.S. Kim, et al., First ATLAS Domestic Standard Problem (DSP-01) for the Code Assessment, Nuclear Enineerin and Technoloy, 43 (1), pp , (2011). 31

44 [9] U.S. Nuclear Reulatory Commission, RELAP5/MOD3.3 Code Manual, NUREG/CR-5535/Rev1, (2001). [10] U.S. Nuclear Reulatory Commission, TRACE V 5.0 Theory Manual, (2010). [11] Korea Atomic Enery Research Institute, MARS Code Manual Volume I: Code Structure, System Models, and Solution Methods, (2009). [12] K.H. Kan, et al., ATLAS Facility and Instrumentation Description Report, KAERI/TR-3779/2009, (2009). [13] I.C. Chu, et al., Development of Passive Flow Controllin Safety Injection Tank for APR1400, Nuclear Enineerin and Desin, 238 (1), pp , (2008). [14] M.J. Thurood, et al., COBRA/TRAC - A Thermal-Hydraulics Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems, NUREG/CR-3046, PNL-4385, (1982). [14] Korea Atomic Enery Research Institute, ATLAS Domestic Standard Problem (DSP) Specifications, (2009). [15] H.R. Ko and T. Kim, "Analysis of a DVI Line Break Accident of the ATLAS Facility," Proceedin of Winter Meetin of American Nuclear Society, (2013). [16] S. G. Bankoff, et al., Countercurrent Flow of Air/Water and Steam/Water throuh a Horizontal Perforated Plate, International Journal of Heat and Mass Transfer, 24, pp , (1981). [17] R. E. Henry and H. K. Fauske, The Two-Phase Critical Flow of One- Component Mixtures in Nozzles, Orifices, and Short Tubes, Transactions of ASME, Journal of Heat Transfer, 93, pp , (1971). 32

45 [18] Ki-Youn Choi and et. al., Quick-Look Data Report on SB-DVI-08, ATLAS Project Experimental Data/Information Transfer, ATLAS-QLR-SB-DVI-08/Rev0, pp. 44, (2009). [19] K. Kan and et. al., "Detailed Description Report of ATLAS Facility and Instrumentation, KAERI/TR-4316/2011," Korea Atomic Enery Research Institute, pp. 179, (2011). [20] Korea Electric Power Co. and Korea Hydro & Nuclear Power Co., APR1400 Desin Control Document Tier 2, Chapter 15 Transient and Accident Analyses, APR1400-K-X-FS NP/Rev0, pp , (2014). [21] Korea Electric Power Co. and Korea Hydro & Nuclear Power Co., APR1400 Desin Control Document Tier 2, Chapter 15 Transient and Accident Analyses, APR1400-K-X-FS NP/Rev0, pp , (2014). 33

46 34

47 Appendix A CCFL MODEL APPENDIX A.1 COFFICIENTS OF CCFL CORRELATION Counter-current flow limitation phenomena describes inflow liquid limitation due to upward vapor flow which is present at certain structures such as upper core tie plate, downcomer annulus, steam enerator tube support plates, and the tube sheet entrance in the steam enerator inlet plenum. Table A.1 Nomenclature of CCFL Calculation Nomenclature c m Hf H j jf v vf α αf r w L Dh Gas Intercept Constant Neative Slope Dimensionless Fluid Flux Dimensionless Gas Flux Gas Superficial Velocity Fluid Superficial Velocity Gas Velocity Fluid Velocity Gas Volume Fraction Fluid Volume Fraction Density Lenth Scale Laplace Capillary Constant Hydraulic Diameter In MARS-KS, the CCFL phenomena is modeled by usin the Bankoff correlation (eq.1), and transformed into Wallis form if lenth scale (eq.4) is altered by user input coefficient (β=0 in eq. 4), which results lenth scale equal to hydraulic 35

48 36 diameter (eq.5). In this calculation, the aim is to find the scalin relations for the neative slope, m and as intercept coefficient, c at the upper core tie plate reion. In order to perform this, Bankoff equation is transformed in Wallis form and rearraned to a more simplified form (eq.10). APPENDIX A.2 CALCULATION PROCEDURE c H m H f 1/ 2 1/ 2 (1) 2 1/ f w j H (2) 2 1/ f f f f w j H (3) L D w h 1 (4) For Wallis, h 0 w D (5) 2 1/ 2 1/ 2 1/ 2 1/ 2 1/ 2 1/ f h f f f h f D j m D j c H m H (6) c D v m D v f h f f f f h 2 1/ 2 1/ 2 1/ 2 1/ (7)

49 D h f 1/ 4 X (8) f f Dh f f 1/ 4 X (9) 1 / 2 1/ 2 (8), (9) in (7) v X m v X c (10) f f The notation in equation is decided to be used in order to use perform scalin easily. Since ATLAS had desined with same pressure and temperature conditions as APR1400, the constituents of X, and X f terms should be similar for both ATLAS and APR1400 calculations. In addition, in order to preserve the heat transfer coefficient between ATLAS and prototype planet, namely APR1400, hydraulic diameter was kept same at the fuel reion, and at top of fuel. Includin the 1/1.414 velocity scalin ratio, and considerin all the characteristics relations mentioned above, the scalin relation of the CFFL correlation s coefficients arraned as follows: r 1/ 2 v r X r m r 1/ 2 v f r X f r C (11) Where: rv v v. APR1400. ATLAS v. APR v. APR (11a) rv f v v f. APR1400 f. ATLAS v f. APR v f. APR (11b) 37

50 rx X X. APR1400. ATLAS 1 (11c) rx f X X f. APR1400 f. ATLAS 1 (11d) m APR1400 rm matlas (11e) c APR1400 rc catlas (11f) Rewritin the equation (11) and includin the (11a, b f): / 2 r / 2 m r C (12) r m r C (13) In order to calculate the coefficients of CCFL correlation for APR1400 at the top of the core, it is essential to perform an ATLAS simulation and find the proper coefficients that show ood areement with LSC. After that, the CCFL coefficients that were obtained from ATLAS simulation can be used to derive the CCFL coefficients APR1400. In order to find one of the CCFL coefficients, an arbitrary coefficient must be chosen and the other one can be calculated by usin equation

51 Understandin the relationship between as intercept constant and neative slope, helps to determine the coefficient and its value. Fiure A.2 Fluid vs. Gas Flux for Different CCFL Coefficients The relationship between fluid and as flux for various CCFL coefficients is depicted in Fiure 7-1. As it can be seen from the fiure, maximum fluid flow can be obtained by settin the as intercept coefficient maximum (e.. c=1) where neative slope tends to et closer to zero. The behavior can be seen on the reen line in Fiure 7-1. On the contrary, neatively slope coeffient must be set to maximum (e.. m=1) in order to have less inflow fluid which is shown as in the dashed black line. By considerin the proper LSC behavior at ATLAS, a previous study performed by H.R. Ko. et. al. suested that the CCFL coefficients should be 1.0 and 0.82 for the as intercept coeffient and neative slope, respectively. In case of APR1400, both coefficients were fixed at a value of 1.0 and the same steps followed to calculate the proper coefficients. It was found out that 1.0 and 0.45 are the ideals 39

52 values for as intercept and neative slope coefficients, respectively. However, it was decided to employ c = 1.0 and m = 0.4 for as intercept and neative slope coefficients, due to existence eometrical distortions and small differences in pressure/temperature between ATLAS and APR

53 Appendix B NODALIZATION OF MODELS APPENDIX B.1 APR1400 NODALIZATION Fiure B.1 APR1400 Nodalization 41

54 APPENDIX B.2 ATLAS NODALIZATION Fiure B.2 ATLAS Nodalization 42