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NUREG/IA-504, International Agreement Report Assessment of Trace 5.0 Against ROSA-2 Test 3 Counterpart Test to Pkl
ML19064B308
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Issue date: 03/31/2019
From: Gallardo S, Lorduy M, Querol A, Kirk Tien, Verdu G
NRC/RES, Universitat Politecnica de Valencia
To:
Meyd, Donald
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NUREG/IA-0504 International Agreement Report Assessment of TRACE 5.0 Aga in st ROSA-2 Test 3 Counterp art Tes t to PKL Prepared by: S.Gallardo, A. Querol, M. Lorduy,G.VerduUniv ersitat Politcnica de Valncia Instituto Universitari o de Segurid ad Industrial , Radiofísica y Medioambiental C amí de V era s/n 46022 Valencia, SPAIN Kirk Tien, NRC Project Manager Division of Systems Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-00 01 Manuscript Completed:

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NUREG/IA-0504 International Agreement Report Assessment of TRACE 5.0 Aga in st ROSA-2 Test 3 Counterp art Tes t to PKL Prepared by:

S.Gallardo, A. Querol, M. Lorduy,G.VerduUniv ersitat Politcnica de Valncia Instituto Universitari o de Segurid ad Industrial , Radiofísica y Medioambiental C amí de V era s/n 46022 Valencia, SPAIN Kirk Tien, NRC Project Manager Division of Systems Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0 001 Manuscript Completed:

April 2018 Date Published: March 2019 Published by U.S. Nuclear Regulatory Commission

iii ABSTRACTThe purpose of this work is to overview the results obtained by the simulation of the Counterpart Test 3 PKL-ROSA (SB-HL-18 in JAEA) in the Large Scale Test Facility (LSTF) using the thermal-hydraulic code TRACE5 patch 2. This experiment simulates a PWR hot leg Small Break Loss-Of-Coolant Accident (SBLOCA). One of the main objectives of this test is to establish a relationship between the Core Exit Temperature (CET) measured by the thermocouples and the fuel rod surface temperature (Peak Cladding Temperature, PCT). The core exit thermocouples are used as an important indicator to start an accident management (AM) operator action by detecting core temperatur e excursion during reactor accidents. Test 3 provides experimental data to study the relation between CET and PCT and the time delay existing between them. A detailed model of the LSTF and the control logic of the Test 3 have been simulated using TRACE5 patch 2. The main thermal hydraulic variables obtained with TRACE5 have been compared with experimental data. In general, the simulation results are able to reproduce the experimental behavio

r.

v FOREWORD Thermalhydraulic studies play a key role in nuclear safety. Important areas where the significance and relevance of TH knowledge, data bases, methods and tools maintain an essential prominence are among others: assessment of plant modifications (e.g., Technical Specifications, power uprates, etc.);analysis of actual transients, incidents and/or start-up tests;development and verification of Emergency Operating Procedures;providing some elements for the Probabilistic Safety Assessments (e.g., success criteria and available time for manual actions, and sequence delineation) and its applicationswithin the risk informed regulation framework;training personnel (e.g., full scope and engineering simulators); and/orassessment of new designs.For that reason, the history of the involvement in Thermalhydraulics of CSN, nuclear Spanish talks about Spain participation in LOFT-OCDE and ICAP Programs took place. Since then, CSN has paved a long way through several periods of CAMP programs, promoting coordinated joint efforts with Spanish organizations within different periods of associated national programs (i.e., CAMP-Espana). From the CSN perspective, we have largely achieved the objectives. Models of our plants are in place, and an infrastructure of national TH experts, models, complementary tools, as well as an ample set of applications, have been created. The main task now is to maintain the expertise, to consolidate it and to update the experience. We at the CSN are aware on the need of maintaining key infrastructures and expertise, and see CAMP program as a good and well consolidated example of international collaborative action implementing recommendations on this issue. Many experimental facilities have -hydraulic database (both separated and integral effect tests). However there is a continuous need for additional experimental work and code development and verification, in areas where no emphasis have been made along the past. On the basis of the SESAR/FAP 1 Nuclear 2001) Support Facilities for Existing and Advanced Reactors , CSNI is promoting since the beginning of this century several collaborative international actions in the area of experimental TH research. These reports presented some findings and recommendations to the CSNI, to sustain an adequate level of research, identifying a number of experimental facilities and programmes of potential interest for present or future international collaboration within the nuclear safety community during the coming decade. The different series of PKL, ROSA and ATLAS projects are under these premises. CSN, as Spanish representative in CSNI, is involved in some of these research activities, helping in this international support of facilities and in the establishment of a large network of international collaborations. In the TH framework, most of these actions are either covering not 1

vi enough investigated safety issues and phenomena (e.g., boron dilution, low power and shutdown conditions, beyond design accidents), or enlarging code validation and qualification data bases incorporating new information (e.g., multi-dimensional aspects, non-condensable gas effects, passive components). This NUREG/IA report is part of the Spanish contribution to CAMP focused on: Analysis, simulation and investigation of specific safety aspects of PKL2/OECD andROSA2/OECD experiments.Analysis of applicability and/or extension of the results and knowledge acquired in these projects to the safety, operation or availability of the Spanish nuclear power plants.Both objectives are carried out by simulating the experiments and conducting the plant application with the last available versions of NRC TH codes (RELAP5 and/or TRACE). On the whole, CSN is seeking to assure and to maintain the capability of the national groups with experience in the thermalhydraulics analysis of accidents in the Spanish nuclear power plants. Nuclear safety needs have not decreased as the nuclear share of the nations grid is expected to be maintained if not increased during next years, with new plants in some countries, but also with older plants of higher power in most of the countries. This is the challenge that will require new ideas and a continued effort. Rosario Velasco García, CSN Vice-president Nuclear Safety Council (CSN) of Spain vii TABLE OF CONTENTSABSTRACT ............................................................................................................................... iii FOREWORDv .ix LIST OF TABLES...........xi EXECUTIVE

SUMMARY

...x iii ACKNOWLEDGMENTS ........................................................................................................... xv ABBREVIATIONS AND ACRONYMS ....................................................................................

xv ii 1 INTRODUCTION ................................................................................................................ 1 2 LSTF FACILITY DESCRIPTION ......................................................................................... 3 3 TRANSIENT DESCRIPTION .............................................................................................. 5 4 TRACE5 MODEL OF LSTF ................................................................................................ 9 5 RESULTS AND DISCUSSION ..........................................................................................

13 5.1 Steady-State ...........................................................................................................

13 5.2 Transient .................................................................................................................

14 5.3 System Pressures ...................................................................................................

15 5.4 Break .......................................................................................................................

16 5.5 Primary L oop Mass Flows .......................................................................................

17 5.6 Vessel Collapsed Liquid Levels ...............................................................................

19 5.7 Maximum Fuel Rod Surface and Core Exit Temperatures .......................................

22 5.8 Hot and Cold Legs Liquid Levels .............................................................................

23 5.9 Emergency Core Cooling Systems Mass Flow Rates ..............................................

26 5.10 Core Power .............................................................................................................

28 5.11 Voi d Fraction ...........................................................................................................

28 6 CONCLUSIONS ................................................................................................................

35 7 REFERENCES ..................................................................................................................

37

ix LIST OF FIGURES Figure 1 Schematic View of the LSTF Facility ........................................................................ 3 Figure 2 Model Nodalization ..................................................................................................

10 Figure 3 Primary and Secondary Pressures ..........................................................................

16 Figure 4 Break Mass Flow Rate ............................................................................................

17 Figure 5 Primary Loop A Mass Flow Rate. ............................................................................

18 Figure 6 Primary Loop B Mass Flow Rate. ............................................................................

18 Figure 7 Core Collapsed Liquid Level....................................................................................

19 Figure 8 Upper Plenum Collapsed Liquid Level .....................................................................

20 Figure 9 Downcomer Collapsed Liquid Level ........................................................................

21 Figure 10 Fuel Rod Surface Temperatures at Different Axial Positions. ..................................

22 Figure 11 Maximum Fuel R od Surface Temperature vs Core Exit Temperature.

.....................

23 Figure 12 Collapsed Liquid Level in the Hot Leg A. ................................................................

.24 Figure 13 Collapsed Liquid Level in the Hot Leg B24 Figure 14 Collapsed Liquid Level in the Cold Leg A.25 Figure 15 Collapsed Liquid Level in the Cold Leg B..25 Figure 16 High Pressure Injection System Mass Flow Rate.. ..................................................

26 Figure 17 Accumulator Injection System Mass Flow Rate...27 Figure 18 Low Pressure Injection System Mass Flow Rate.27 Figure 19 Core Power .............................................................................................................

28 Figure 20 Void Fraction in the LSTF at 0 s ..............................................................................

29 Figure 21 Void Fraction in the LSTF when PCT Reaches 750 K and HPI Starts30 Figure 22 Void Fraction in the LSTF when Primary Pressure = 5 MPa. ...................................

31 Figure 23 Void Fraction in the LSTF when Second PCT Excursion is Produced

.....................

32 Figure 24 Void Fraction in the LSTF at the End of the Transient33

xi LIST OF TABLES Table 1 Control Logic and Sequence of Major Events in the Experiment ................................ 6 Table 2 Core Protection Logic ................................................................................................. 7 Table 3 Vessel Nodalization. ................................................................................................... 9 Table 4 Steady-State Comparison between Experimental and Simulated Values13 Table 5 Sequence of Events Comparison between Experiment and TRACE5.. .....................14

xiii EXECUTIVE

SUMMARY

The purpose of this work is to test the capability of the thermal hydraulic code TRACE5 in the simulation of a Hot Leg Small Break LOCA (SBLOCA) in the frame of the OECD/NEA ROSA-2 Project. The main objective of this project is providing experimental data in the Large Scale Test Facility (LSTF) for assessment of thermal hydraulic computer codes. The transient considered in this work, Test 3, reproduces a PWR 1.5% hot leg SBLOCA with an assumption of total failure of High Pressure Injection (HPI) system. Test 3 was designed as counterpart test between PKL and LSTF test facilities. One of the purposes of this test is to clarify the relation between the Core Exit Temperature (CET) measured by thermocouples and the Peak Cladding Temperature (PCT) at high and low-pressure conditions corresponding to the pressure range of LSTF and PKL facilities during a hot leg SBLOCA. TRACE5 patch 2 has been used to model the LSTF and the control logic of Test 3. The model considers the high-pressure phase, the conditioning phase and the low-pressure phase performed in the LSTF experiment.

The behavior of the Pressure Vessel is analyzed, measuring the active core, upper plenum, upper head and downcomer liquid levels. Results of the simulation with TRACE5 are compared with the experimental measurements in several graphs, including primary and secondary pressures, break mass flow rate, primary mass flow rates, and collapsed liquid levels (hot leg, steam generators U-tubes, etc.).

xv ACKNOWLEDGMENTS This paper contains findings that were produced within the OECD-NEA ROSA-2 Project. The authors are grateful to the Management Board of the ROSA Project for their consent to this publication, and thank the Spanish Nuclear Regulatory Body (CSN) for the technical and financial support under the agreement STN/1388/05/748.

xvii ABBREVIATIONS AND ACRONYMS AFW Auxiliary Feedwater AIS Accumulator Injection System AM Accident Management BE Best Estimate CAMP Code Assessment and Management Program CET Core Exit Temperature CPU Central Processing Unit CRGT Control Rod Guide Tubes CSN Nuclear Safety Council, Spain DBE Design Basis Event ECCS Emergency Core Cooling System HPI High Pressure Injection IBLOCA Intermediate Break Loss-Of-Coolant Accident JAEA Japan Atomic Energy Agency JAERI Japan Atomic Energy Research Institute JC Jet Condenser LOCA Loss-Of-Coolant Accident LPI Low Pressure Injection LSTF Large Scale Test Facility MFW Main Feedwater MSIV Main Steam Isolation Valve NPP Nuclear Power Plant NRC U.S. Nuclear Regulatory Commission NV Normalized to the Steady State Value PA Auxiliary Feedwater Pump PCT Peak Cladding Temperature PF Feedwater Pump PGIT Primary Gravity Injection Tank PJ High Pressure Charging Pump PL High Pressure Injection Pump PV Pressure Vessel PWR Pressurized Water Reactor PZR Pressurizer RHR Residual Heat Removal RV Relief Valve SBLOCA Small Break Loss-Of-Coolant Accident SG Steam Generator SI Safety Injection SNAP Symbolic Nuclear Analysis Package SRV Safety Relief Valve ST Storage Tank

1 1 INTRODUCTION The purpose of this work is to describe the most relevant results achieved by using the thermal hydraulic code TRACE5 patch 2 [1 , 2] to simulate the SBLOCA transient defined in Test 3 within the OECD/NEA ROSA-2 Project (SB-HL-18 in JAEA) [3]. This transient was performed in the Large Scale Test Facility (LSTF) [4], which simulates a PWR reactor, Westinghouse type, of four loops and 3423 MW of thermal power, scaled to 1/48 in volume and two loops. Thermocouples are common ly used as an important indicator to start an Accident Management (AM) action by detecting the Core Exit Temperature (CET) excursion during reactor accidents. However, in some tests , a time delay between the detection of superheated steam by thermocouples and the CET excursion is observed [5, 6 , 7]. The CET reliability to detect core uncover during a SBLOCA is considered as one of the most important safety concerns studied in the ROSA-2 Project. Test 3 was designed as counterpart test between the PKL and LSTF test facilities. One of the purposes of this test is to clarify the relation between the CET measured by thermocouples and the fuel rod surface temperature (PCT) at high and low-pressure conditions corresponding to the pressure range of LSTF and PKL facilities during a hot leg SBLOCA. Another goal of this counterpart test is to study two large integral test facilities with different designs to draw technical findings useful for solving scaling problems.

3 2 LSTF FACILITY DESCRIPTION In this section, a brief description of the LSTF facility [10] is presented. LSTF simulates a PWR reactor, Westinghouse type, of four loops and 3423 MW of thermal power. The facility is electrically heated, scaled 1:1 in height and 1:48 in flow areas and volumes, with exception of the loops, which are defined by a scaling factor of 1:24 in flow areas and volumes. Figure 1 shows the scheme of the LSTF facility. As it can be seen, the primary coolant system consists of the Pressure Vessel (PV) and two symmetrical primary loops: loop A with the pressurizer (PZR) and loop B. Figure 1 Schematic View of the LSTF Facility Each loop contains a primary Coolant Pump (PC) and a Steam Generator (SG). The secondary-coolant system consists of the jet condenser (JC), the Feedwater Pump (PF), the Auxiliary Feedwater Pumps (PA) and related piping system in addition to two SG secondary systems. The Pressure Vessel (PV) is composed of an upper head above the upper core support plate, the upper plenum between the upper core support plate and the upper core plate, the core, the lower plenum and the downcomer annulus region surrounding the core and upper plenum. LSTF vessel has 8 upper head spray nozzles (of 3.4 mm inner-diameter), and 8 Control Rod Guide Tubes (CRGTs) characterize the flow path between the upper head and upper plenum. The maximum core power of the LSTF is limited to 10 MW which corresponds to 14% of the volumetrically scaled PWR core power and is sufficiently capable to simulate PWR decay heat power after the reactor scram. Regarding the SGs, each of them contains 141 U-tubes which can be classified into separate groups depending on their length. The U-tubes have an inner diameter of 19.6 mm and an outer diameter of 25.4 mm (with 2.9 mm wall thickness). On the other hand, vessel, plenum and riser of steam generators have a height of 19.840, 1.183 and 17.827 m, respectively. The downcomer is 14.101 m in height.

5 3 TRANSIENT DESCRIPTION Test 3 reproduces a PWR hot leg Small Break LOCA, which flow area corresponds to 1.5% of reference PWR cold leg area, with High Pressure Injection (HPI) into the PV upper plenum [3].

The transient performed in LST F is divided in to three phases; high-pressure transient phase, conditioning phase, and low-pressure transient phase to meet the PKL pressure. The complete control logic of the transient is listed in Table 1. The transient starts at time 0 with opening the break valve in the hot leg of loop without pressurizer and increasing the rotational speed of the coolant pumps. Few seconds afterwards, the scram signal is generated. This signal produces the initiation of the core power decay curve. In addition, the scram signal produces the initiation of the primary coolant pumps coast, the turbine trip, the closure of the Main Steam Isolation Valves (MSIV) and the termination of the Main Feedwater (MFW).

To protect the facility, the LSTF Core Protection System automatically decreases the core power when the maximum fuel rod surface temperature reaches 958 K, as it can be seen in Table 2. Immediately after the maximum fuel rod surface temperature reaches 750 K, the HPI system injects coolant into the PV upper plenum to avoid subcooled water layer being formed at the PV bottom. This phase is terminated when the primary pressure decreases to 5 MPa and the break valve is temporarily closed. In the conditioning phase, the core power is manually changed to a constant value (1.16 MW). The primary mass inventory is recovered by the continuous HPI injection into the PV upper plenum. When the hot leg liquid level recovers up to the middle level, the HPI is terminated. The Relief Valves (RVs) are fully opened in both SGs for depressurization. The Auxiliary Feedwater (AFW) is then injected in to both SGs to avoid significant liquid level drop. When the primary pressure decreases to 3.9 MPa, the RVs are closed and AFW is terminated in both SGs. This phase is finished when the primary pressure reaches 4.5 MPa. In the low-pressure phase, the break valve is again opened. Due to the coolant loss through the break, the core uncover is produced. Immediately after the CET reaches 623 K, the SG secondary-side depressurization is initiated by fully opening the RVs at both SG as an AM action. The AFW is also injected in to both SGs. The Accumulator Injection system (AIS) is initiated when the primary pressure reaches 2.6 MPa.

The Low Pressure Injection (LPI) system is actuated when the PV lower plenum pressure reaches 1 MPa. The transient is terminated when continuous core cooling by the LPI system is confirmed.

6 Table 1 Control Logic and Sequence of Major Events in the Experiment Break. Time zero High-pressure transient phase Generation of scram signal.

Primary pressure = 12.97 MPa Pressurizer (PZR) heater off.

Generation of scram signal or PZR liquid level below 2.3 m Initiation of core power decay curve.

Generation of scram signal.

Initiation of Primary Coolant Pump coastdown. Generation of scram signal.

Turbine trip (closure of steam generators Main Steam Isolation Valve s). Generation of scram signal.

Closure of steam generators Main Steam Isolation Valves.

Generation of scram signal.

Termination of steam generators Main Feedwater.

Generation of scram signal.

Generation of Safety Injection (SI) signal. Primary pressure = 12.27 MPa Initiation of High Pressure Injection system (HPI) into the pressure vessel upper plenum.

Maximum fuel rod surface temperature = 750 K

. Initiation of steam generators secondary-side depressurization by fully opening of Relief Valves in both loops as AM action.

Maximum core exit temperature = 623 K Low-pressure transient phase Initiation of Auxiliary Feed Water in both loops Initiation of AM action.

Initiation of Accumulator Injection system in both loops Primary pressure = 2.6 MPa Termination of Accumulator Injection system in both loops Primary pressure =1.2 MPa Initiation of Low Pressure Injection system in both loops PV lower plenum pressure = 1 MPa 7 Table 2 Core Protection Logic Core power to Maximum fuel rod surface temperature (K) 70% 958 K 35% 961 K 13% 966 K 5% 977 K 0% (core power trip) 1003 K

9 4 TRACE5 MODEL OF LSTF LSTF has been modeled with 97 hydraulic components (11 BREAKs, 12 FILLs, 25 PIPEs, 2 PUMPs, 1 PRIZER, 26 TEEs, 19 VALVEs and 1 VESSEL). Figure 2 shows the nodalization of the model using the Symbolic Nuclear Analysis Package software (SNAP) [9]. Primary side comprises cold and hot legs, pumps, loop seals, a pressurizer in loop A, the ECCS which includes AIS, HPI and LPI systems, the U-tubes of both SG and the PV. On the other hand, secondary side includes steam separators, risers, downcomers, Safety Relief Valves , Main Steam Isolation Valves and FILLs to provide Main Feedwater and Auxiliary Feedwater. The pressure vessel has been modelled using a 3DVESSEL component. The VESSEL nodalization consists of 20 axial levels, 4 radial rings and 10 azimuthal sectors. For each axial level, volume and effective flow area fractions have been set according to technical specifications provided by the organization [4]. Table 3 shows the vessel nodalization in the axial direction. Table 3 Vessel Nodalization Levels Parts of the vessel 1-2Lower plenum 3-11Core 12 Upper core plate 13-16Upper plenum 17 Upper core support plate 18-20Upper head The 3D-VESSEL is connected to different 1D components: 8 Control Rod Guide Tubes (CRGT), hot leg A and B (level 15), cold leg A and B (level 15) and a bypass channel (level 14). The CRGTs have been simulated by PIPE components, connecting levels 14 and 19 and allowing the flow between upper head and upper plenum. 30 HTSTR components simulate the fuel assemblies in the active core. A POWER component manages the power supplied by each HTSTR to the 3D-VESSEL. Fuel elements (1008 in total) were distributed into the 3 rings: 154 elements in ring 1, 356 in ring 2 and 498 in ring 3. In both axial and radial direction, different peaking factors have been considered. The power ratio in the axial direction presents a peaking factor of 1.495, while the radial power profile is divided into three power zones using the first three radial rings. Depending on the radial ring, different peaking factors have been considered (0.66 in ring 1, 1.51 in ring 2 and 1.0 in ring 3). The number of fuel rod components associated with each heat structure has been determined from the technical documentation given, considering the distribution of fuel rod elements in the vessel.

10 A detailed model of SG (geometry and thermal features) has been developed. Boiler and downcomer components of secondary-side have been modeled by TEE components. U-tubes have been classified into three groups according to each average length. Figure 2 Model Nodalization The steam separator can be simulated in TRACE5 setting a friction coefficient (FRIC) greater than 10 22 at a determined cell edge, allowing only gas phase to flow through the cell interface. Heat transfer between primary and secondary sides has been performed using HTSTR components. Cylindrical-shape geometry has been used to best fit heat transmission. Inner and outer surface boundary conditions for each axial level have been set to couple HTSTR component to hydro components (primary and secondary fluids). Different models varying the number of U-tube groups were tested (1, 3 and 6 groups). It was found that the results do not apparently change, using these models. Heat losses to the environment have been considered in the secondary-side walls.

U-tubesLoop B U-tube sPressurizer Accumulators LPI LPI Secondary side A Secondary side B Loop A Break Valve HPI during Test 3 H PI H PI 11 Choke model predicts for a given cell the conditions for which choked flow is expected to occur, providing three different models: subcooled-liquid, two-phase and single-phase vapor model. TRACE5 patch 2 code allows to choose the subcooled-liquid and two-phase coefficients. In this case, the default values (1.0) have been selected. The break has been simulated by means of a VALVE component connected to a BREAK component to establish the boundary conditions.

13 5 RESULTS AND DISCUSSION 5.1 Steady- tate Steady-state conditions achieved in the simulation w ere in reasonable agreement with the experimental values. It can be seen i n Tabl e 4, where the relative error s (%) between experimental and simulated result s for different item s are listed.

To achieve the steady state conditions, the duration of simulation was stated t o 1000 s. Tabl e 4 Steady-St ate Condition Comparison between Experimental an d Simulated Values Item (Loop with PZR)

Relative Error (%) Core Power 0.00 Hot leg Fluid Temperature 0.10 Cold leg Fluid Temperature 0.27 Mass Flow Rate 3.68 Pressurizer Pressure 0.06 Pressurizer Liquid Level 3.60 Accumulator System Pressure

-0.38Accumulator System Temperature

-0.56SG Secondary

-side Pressure 0.96 SG Secondary

-side Liquid Level 5.67 Steam Flow Rate 3.86 Main Feedwater Flow Rate 3.40 Main Feedwater Temperature 0.06 Auxiliary Feedwater Temperature

-0.13 14 5.2 Transient Tabl e 5 list s the chronological sequence of the transient event s and the comparison between the experiment and TRACE. Tabl e 5 Chronological Sequence of Events Comparison between Experiment and TRACE5 Event Experiment Time (s) TRACE5 Time (s) Break valve open 0 0 Scram signal 29 32 Closure of steam generators (SG) Main Steam Isolation Valves 32 33 Initiation of coastdown of primary coolant pumps 33 35 Termination of SG main feedwater 34 33 SI signal 37 43 Initiation of core power decay 50 47 Primary coolant pumps stop 281 280 Primary pressure lower than secondary side pressure 1310 1317 T he increasing in fuel rod surface temperature starts 1595 1680 Maximum fuel rod surface temperature reached 750 K 1840 1796 Initiation of High Pressure Injection (HPI) into the PV upper plenum 1844 1796 Break valve closure 2172 2113 Manual change of core power to a constant value 2215 2500 Termination of HPI system into the PV upper plenum 2852 2880 Initiation of SG secondary

-side depressurization by fully opening Relief Valves (RVs) in both loops 2880 2893 Initiation of Auxiliary Feed Water (AFW) in both loops 2900 2893 Termination of SG secondary

-side depressurization 3028 2959 Termination of AFW in both loops 3055 2959 Break valve open again 3323 3303 Start of the increasing in fuel rod surface temperature 3983 4203 15 Primary pressure becomes lower than SG secondary side pressure 4105 4250 Maximum core exit temperature = 623 K 4390 4507 Initiation of SG secondary

-side depressurization by fully opening RVs in both loops as AM action 4394 4519 Initiation of AFW in both loops 4410 4519 Maximum fuel rod surface temperature 4413 4545 Initiation of Accumulator Injection system (AIS) in both loops 4500 4590 Termination of AIS in both loops 4829 4786 Initiation of Low Pressure Injection (LPI) system in both loops 5003 4830 5.3 System Pressures A comparison between primary an d secondary pressures i s presented i n Figure 3. In the high-pressure phase (until 2100 s), the primary pressure start s t o decrease when the brea k valve is opened due t o the coolant discharged through the break. The scram signal i s generated when the primary pressure i s low er than 12.97 MPa. The SI signal is generated when the primary pressure reache s 12.27 MPa. The generation of the scr am signal cau ses the closure of SG MSIVs an d the beginning of the primary coolant pumps coastdow

n. The SG secondary-side pressur e rapidly increases after the closure of MSIVs. From thi s moment on, the SG secondary-side pressure start s to oscill ate by opening and closing the RV s of both SGs. The primary pressure becomes low er tha n the SG seconda ry-side pressure at about 1 250 s soon after the brea k flow turns into single-phase steam (Figure 4) and decreases until 5 MPa (at 2170 s), when the brea k valve is closed.

I n the conditioning phase (until 33 00 s), the primary pressure increases once up to 6.5 MP a and decreases to 4 MPa following the SG secondary-side depressurizatio

n. The primary and secondary pressures increase again up to about 4.5 MPa after the termination of the SG secondaryside depressurization. The low-pressure pha se start s by opening the break valve again (at 3300 s). The primary pressure becomes low er than the SG pressure at about 4100 s, slightly aft er the core boil-off is produced. Immediately after the CET reaches 623 K (Figure 10), the SG secondaryside depressurization is initiated. Then, the primary pressure decreases following the SG pressure and activates the Accumulator Injection and the Low Pressur e Injection systems.

I n general, both primary and secondary-side pres sures ar e successfully reproduced by TRACE5 in the whole transient. However, some discrepanci es are observed regarding the pressure drop slopes, which ar e different in all the cases.

16 Figure 3 Primary and Secondary Pressures 5.4 Break Figure 4 show s the mass flow rate throug h the bre ak. I n the high-pressure phase, the brea k flow rate decrease s when the brea k flow turns from single-phase liqui d t o two-phase flow (200 s). At 1250 s, the brea k flow tur ns t o single-phase vapor. To adjust the break ma ss flow rate with TRACE 5, a sensitivity analysi s varying the discharge coefficient of the Choked Flow model was performed. I n the results show n, the discharge coefficients have been fixed to 1.0. During the interval between 2 50 and 1 000 s, the mass flow rate through the break obtained with TRACE5 is higher than the experimental values. However, the changes from liquid to two-phase and fr om two-phase to one phase vapor are reproduced at similar time. LOW PRESSURE PHASE CONDITIONING PHASE Scram signal MSIVs closure RV cycling open/close Break valve closure 1 st depressurization 2 nd depressurization HPI LPI starts HIGH PRESSURE PHASE AIS starts

17 Figure 4 Break Ma ss Flow Rate 5.5 Primary Loop Ma ss Flows Figures 5 and 6 show the primary mass flow rate in both loops. During the first second s of the transient, the primary ma ss flow increases du e t o the higher angul ar speed of the pumps. Then, the primary mass flow rate in both loops decreases according to the pump coastdown. The main discrepanci es are found during the accumulator water injection. At 4500 s, the experimental primary ma ss flow rate in both loops are higher tha n the simulation results du e to the accumulator water entrance i n the cold legs. Thi s effect ha s not bee n properly predicted by TRACE5. LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

18 Figure 5 Primary Loop A M ass Flow Rate Figure 6 Primary Loop B M ass Flow Rate LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

19 5.6 Vessel Collapsed Liquid Levels The Figur es 7 , 8 and 9 show a comparison between the collapsed liquid level s in the core, upper plenum an d downcomer, respectively, obtaine d for both experimental an d simulation results. I n the core liquid level, a significant drop start s due to the boil-off at about 1550 s, when the upper plenum i s emptied. The core uncovering takes plac e after the primary pressure becomes low er than the secondary pressure. The collapsed liquid level continues to drop to about 1/3 of the active core len gth until 1870 s, even after the initiation of the high-pressure coolant injection i nto the PV upper plenum.

I n the conditioning phase, the liquid level in the core and the upper plenum i s recov ered du e to the high-pressure coolant injection. When the upp er plenum reache s the middle level (at 2850 s), the HPI finishes. Figure 7 Core Collapsed Liquid Level When the break valve is opened again at 3300s to start the low-pressure phase, the liquid level drops in the upper plenum and the core. The core liquid level begins to drop at about 3900 s, and the core uncovering is produced before the primary pressure becomes lower than the secondary pressure. The core liquid level is recovered following the primary depressurization. LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

20 Figure 8 Upper Plenum Collapsed Liquid Level Regarding the downcomer liquid level in the high pressure phase, it drops gradually up to 1800 s. It is recover ed at 1910 s, after the initiation of the HPI in to the PV upper plenum. In the conditioning phase, the temporary liquid level drop happens as in the upper plenum and the core due to the secondary depressurization. In the low-pressure phase (after 35 00 s), the liquid level drops following the core boil-off. The downcomer liquid level is steeply recover ed at 4530 s due to the accumulator coolant injection. As it can be observed in Figure 9 , some discrepancies appear during the first emptying of the downcomer, which is delayed in the simulation about 500 s. LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

21 Figure 9 Downcomer Collapsed Liquid Level LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

22 5.7 Maximum Fuel Ro d Surface and Core Exit Temperatures Figure 10 show s a comparison between nine axial position s of a fuel rod. A HTSTR from the second radial ring has been chosen. The axial position s correspond to the nine axial level s in which is divided the core. The main discrepancies are observed during the second temperature excursion. TRACE5 doe s not reproduce the maximum value, obtaining a pe ak 100 K lower. In the high-pressure phase, the PCT reache s 750 K at 1840 s and the high pressure injection into the PV upper plenum is initiated. Figure 10 Fuel R od Surface Temperatures at D ifferent Axial Positions In the experiments performed by the NEA Working Group on the Analysis and Management of Accidents [7], a significant difference between CET and PCT was observed in the measurements obtained in all the facilities. Thus, as the CET triggers the AM actions, but the safety variable normally followed in nuclear safety is the PCT, a more detailed study of both variables and the relation between them was suggested. Figure 11 shows the representation CET versus PCT. This figure allows clarifying the relation between CET measured by thermocouples and PCT during a hot leg SBLOCA. This relation allows obtaining the PCT that corresponds to a determined CET. The delay between the PCT and the CET excursions produces that the CET remains constant while the PCT increases to 600 K. However, in TRACE5 both excursions are initiated at the same time. LOW PRESSURE PHAS E CONDITIONING PHASE HIGH PRESSURE PHASE

23 Figure 11 Maximum Fuel Rod Surface Temperature vs Core Exit Temperature 5.8 Hot and Cold Legs Liquid Levels Figures 12 and 13 show the liquid level i n bot h hot legs. A s it can be seen, the liquid level is almost the same in bot h legs (intact and broken loop). In the high-pressure phase, the hot leg fluid becomes saturated at 53 s. The liquid level was kept at around 3/4 to 1/2 of the inner diameter until about 1310 s , when the steam bre ak flow and primary pres sure decreases. The hot leg becomes empty at about 1400 s, when the liquid level in the upper plen um begins to drop. In the conditioning phase, the hot leg liquid level is recover ed at abo ut 2640 s after the core reflooding and reache s the middle level. In the low-pressure phase, the liquid level start s to decrea se just after the brea k valve open. The hot leg becomes empty at abo ut 3600 s. The hot leg liquid level i s recover ed at 4530 s du e to the accumulator injection.

In high-and low-pressure phases, the fluid is kept saturated i n the intact loop and superheated i n the broken loop during the core uncovering. It suggest s that t he steam preferentially flow s tow ards the brea k from the PV upper plenum while stagnat es i n the loo p A, except during the SG depressurization. Figures 14 and 15 show the cold leg liquid levels.

In loop A, the cold leg becomes empty at about 900 s, wherea s in loop B it occur s at 1250 s. However, in the simulation bot h cold legs are empty at the sam e tim e (around 1250 s). During the coolant injection by the accumulator and LPI systems, the liquid level i s recov ered in both loops. However, TRACE does not exactly reproduce the recovering of the liquid level in both cold legs, due to the rapid coolant discharge throug h the break from the accumulation injection.

24 Figure 12 Collapsed Liquid Level in the Hot Leg A Figure 13 Collapsed Liquid Level in the Hot Leg B LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

25 Figure 14 Collapsed Liquid Level in the Cold Leg A Figure 15 Collapsed Liquid Level in the Cold Leg B LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

26 5.9 Emergency Core Cooling Systems Ma ss Flow Rate The HPI into the PV upp er plenum i s activated immediately after the maximum fuel rod surface temperatur e reaches 750 K (Figur e 10). Thi s HPI configuration i s use d t o avoid thermal stratification occurring i n the PV low er plenum. The coolant injection finishes at 2850 s, when the liquid level in the hot leg recov ers the middle level. A s it can be seen in Figure 16, wher e the HPI ma ss flow rates ar e show n, TRACE5 reproduces the HPI injection successfully. Figure 16 High Pressure Injection System Mass Flow Rate Figure 17 shows the coolant injection flow rate from the Accumulator Injection System. As it can be seen, the simulated mass flow rate is lower than the experimental data, due to the different primary pressure drop. This fact could explain the differences in hot and cold legs refill. The LPI system is activated in both loops at 5000 s. In the simulation, this event is advanced as it can be seen in Figure 18. LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

27 Figure 17 Accumulator Injection System Mass Flow Rate Figure 18 Low Pressure Injection System Mass Flow Rate LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

28 5.10 Core Power Figure 19 shows the experimental and the simulated core power curves. The core power starts to decay at 50 s following the core power curve decay. The core power is manually changed to a constant value of 1.16 MW at 2215 s in the experiment and at 2500 s in the simulation. As it can be seen, the simulated core power curve has a good agreement with the experimental curve. Figure 19 Core Power 5.11 Void Fraction Figures 20, 21, 22, 23 and 24 show the void fraction achieved using the LSTF TRACE5 model during the transient, when important events happen. Figure 19 shows the void fraction in the LSTF at the initiation of the test. As it can be seen, primary and secondary sides are full of liquid at this time. Figure 20 shows the void fraction when the PCT reaches 750 K and the HPI injection starts. At this time, the pressurizer is empty, and the liquid is located in the loop seals, accumulators, bottom of the PV and SGs. Figure 21 shows the void fraction when the primary pressure drops to 5 MPa and the break valve is closed. In this moment, the lower plenum of the PV, SG and AIS remain full of liquid, while the loop seals are almost empty. The situation when the second maximum of the PCT is reached is shown in Figure 22. SGs are empty while the loop seals LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE Power (W) 29 have some liquid. Finally, Figure 23 shows the void fraction at the end of the transient. As it can be seen, the SGs are almost empty and the accumulators have not been completely emptied. Figure 20 Void Fraction in the LSTF at 0 s 30 Figure 21 Void Fraction in the LSTF when PCT Reaches 750 K and HPI Starts 31 Figure 22 Void Fraction in the LSTF when Primary Pressure = 5 MPa 32 Figure 23 Void Fraction in the LSTF when Second PCT Excursion is Produced 33 Figure 24 Void Fraction in the LSTF at the End of the Transient

35 6 CONCLUSIONS Results show that TRACE5 can successfully reproduce all the phenomena produced in Test 3 OECD/NEA ROSA-2 Project (SB-HL-18 in JAEA) during the different transient phases: high-pressure, conditioning, and low-pressure. The main variables of the system present a good agreement in comparison to experimental data. System pressures, collapsed liquid levels in the PV , CET and PCT excursions are well reproduced. However, some discrepancies observed in the break mass flow rate could be attributed to the lack of a single-phase vapor coefficient in the choked flow model of TRACE5 patch 2. These differences can affect the mass flow rate through the hot and cold legs. Some discrepancies are also found in the maximum values reached during the second temperature excursions. However, these discrepancies do not affect the relation between the core exit temperature and the fuel rod surface temperature.

37 7 REFERENCES 1.Nuclear Regulatory Commission, Division of Risk Assessment and Special Projects.Office of Nuclear Regulatory Research. U. S Nuclear Regulatory Commission, TRACEMRegulatory Commission, U.S. (2007).

2.Nuclear Regulatory Commission, Division of Risk Assessment and Special Projects,Manual. Volume 1: InputSpecification, Nuclear Regulatory Commission, U. S (2007).

3.Thermohydraulic Safety Research Group, Nuclear Safety Research Center, Final Data Report of ROSA-2/LSTF Test 3 (Counterpart Test to PKL SB-HL-18), Japan AtomicEnergy Agency, JAEA (2010).

4.The ROSA-V Group, ROSA-V Large Scale Test Facility (LSTF) System Description forthe 3rd and 4th Simulated Fuel Assemblies, JAE RI-Tech, Japan (2003).

5.Takeda, T., et al. Quick-look Data Report of OECD/NEA ROSA Project Test 6-1 (1.9%

Pressure Vessel Upper Head Small Break LOCA Experiment). Japan Atomic EnergyAgency, Private Communication. (2006).

6.Freixa J.; Manera A. Analysis of an RPV Upper Head SBLOCA at the ROSA Facility UsingTRACE. Nuclear Engineering and Design, 240, pp. 1779-1788. (2010).

7.Toth, I., Prior, R., Sandervag, O., Umminger, K., Nakamura, H., Muellner, N., Cherubini, TemperatureCommittee on the Safety of Nuclear Installations, (2010).

8.Nakamura, H., Watanabe, T., Takeda, T., Maruyama, Y., Suzuki M., Overview of Recent E fforts through ROSA/LSTF Experiments. Nuclear Engineering and Technology 41(6), pp.

753-764, (2009).9.Nuclear Regulatory Commission and Applied Programming Technology, SymbolicNuclear Analysis Package (SNAP) (2012).

10.Freixa J.; Martnez-Quirogaa V.; Zerkakb O. and Revents F. 2015. Modelling Guidelinesfor Core Exit Temperature Simulations with System Codes. Nuclear Engineering andDesign 286 (2015) 116-129.

NUREG/IA-0504 International Agreement Report Assessment of TRACE 5.0 Aga in st ROSA-2 Test 3 Counterp art Tes t to PKL Prepared by: S.Gallardo, A. Querol, M. Lorduy,G.VerduUniv ersitat Politcnica de Valncia Instituto Universitari o de Segurid ad Industrial , Radiofísica y Medioambiental C amí de V era s/n 46022 Valencia, SPAIN Kirk Tien, NRC Project Manager Division of Systems Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-00 01 Manuscript Completed:

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NUREG/IA-0504 International Agreement Report Assessment of TRACE 5.0 Aga in st ROSA-2 Test 3 Counterp art Tes t to PKL Prepared by:

S.Gallardo, A. Querol, M. Lorduy,G.VerduUniv ersitat Politcnica de Valncia Instituto Universitari o de Segurid ad Industrial , Radiofísica y Medioambiental C amí de V era s/n 46022 Valencia, SPAIN Kirk Tien, NRC Project Manager Division of Systems Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0 001 Manuscript Completed:

April 2018 Date Published: March 2019 Published by U.S. Nuclear Regulatory Commission

iii ABSTRACTThe purpose of this work is to overview the results obtained by the simulation of the Counterpart Test 3 PKL-ROSA (SB-HL-18 in JAEA) in the Large Scale Test Facility (LSTF) using the thermal-hydraulic code TRACE5 patch 2. This experiment simulates a PWR hot leg Small Break Loss-Of-Coolant Accident (SBLOCA). One of the main objectives of this test is to establish a relationship between the Core Exit Temperature (CET) measured by the thermocouples and the fuel rod surface temperature (Peak Cladding Temperature, PCT). The core exit thermocouples are used as an important indicator to start an accident management (AM) operator action by detecting core temperatur e excursion during reactor accidents. Test 3 provides experimental data to study the relation between CET and PCT and the time delay existing between them. A detailed model of the LSTF and the control logic of the Test 3 have been simulated using TRACE5 patch 2. The main thermal hydraulic variables obtained with TRACE5 have been compared with experimental data. In general, the simulation results are able to reproduce the experimental behavio

r.

v FOREWORD Thermalhydraulic studies play a key role in nuclear safety. Important areas where the significance and relevance of TH knowledge, data bases, methods and tools maintain an essential prominence are among others: assessment of plant modifications (e.g., Technical Specifications, power uprates, etc.);analysis of actual transients, incidents and/or start-up tests;development and verification of Emergency Operating Procedures;providing some elements for the Probabilistic Safety Assessments (e.g., success criteria and available time for manual actions, and sequence delineation) and its applicationswithin the risk informed regulation framework;training personnel (e.g., full scope and engineering simulators); and/orassessment of new designs.For that reason, the history of the involvement in Thermalhydraulics of CSN, nuclear Spanish talks about Spain participation in LOFT-OCDE and ICAP Programs took place. Since then, CSN has paved a long way through several periods of CAMP programs, promoting coordinated joint efforts with Spanish organizations within different periods of associated national programs (i.e., CAMP-Espana). From the CSN perspective, we have largely achieved the objectives. Models of our plants are in place, and an infrastructure of national TH experts, models, complementary tools, as well as an ample set of applications, have been created. The main task now is to maintain the expertise, to consolidate it and to update the experience. We at the CSN are aware on the need of maintaining key infrastructures and expertise, and see CAMP program as a good and well consolidated example of international collaborative action implementing recommendations on this issue. Many experimental facilities have -hydraulic database (both separated and integral effect tests). However there is a continuous need for additional experimental work and code development and verification, in areas where no emphasis have been made along the past. On the basis of the SESAR/FAP 1 Nuclear 2001) Support Facilities for Existing and Advanced Reactors , CSNI is promoting since the beginning of this century several collaborative international actions in the area of experimental TH research. These reports presented some findings and recommendations to the CSNI, to sustain an adequate level of research, identifying a number of experimental facilities and programmes of potential interest for present or future international collaboration within the nuclear safety community during the coming decade. The different series of PKL, ROSA and ATLAS projects are under these premises. CSN, as Spanish representative in CSNI, is involved in some of these research activities, helping in this international support of facilities and in the establishment of a large network of international collaborations. In the TH framework, most of these actions are either covering not 1

vi enough investigated safety issues and phenomena (e.g., boron dilution, low power and shutdown conditions, beyond design accidents), or enlarging code validation and qualification data bases incorporating new information (e.g., multi-dimensional aspects, non-condensable gas effects, passive components). This NUREG/IA report is part of the Spanish contribution to CAMP focused on: Analysis, simulation and investigation of specific safety aspects of PKL2/OECD andROSA2/OECD experiments.Analysis of applicability and/or extension of the results and knowledge acquired in these projects to the safety, operation or availability of the Spanish nuclear power plants.Both objectives are carried out by simulating the experiments and conducting the plant application with the last available versions of NRC TH codes (RELAP5 and/or TRACE). On the whole, CSN is seeking to assure and to maintain the capability of the national groups with experience in the thermalhydraulics analysis of accidents in the Spanish nuclear power plants. Nuclear safety needs have not decreased as the nuclear share of the nations grid is expected to be maintained if not increased during next years, with new plants in some countries, but also with older plants of higher power in most of the countries. This is the challenge that will require new ideas and a continued effort. Rosario Velasco García, CSN Vice-president Nuclear Safety Council (CSN) of Spain vii TABLE OF CONTENTSABSTRACT ............................................................................................................................... iii FOREWORDv .ix LIST OF TABLES...........xi EXECUTIVE

SUMMARY

...x iii ACKNOWLEDGMENTS ........................................................................................................... xv ABBREVIATIONS AND ACRONYMS ....................................................................................

xv ii 1 INTRODUCTION ................................................................................................................ 1 2 LSTF FACILITY DESCRIPTION ......................................................................................... 3 3 TRANSIENT DESCRIPTION .............................................................................................. 5 4 TRACE5 MODEL OF LSTF ................................................................................................ 9 5 RESULTS AND DISCUSSION ..........................................................................................

13 5.1 Steady-State ...........................................................................................................

13 5.2 Transient .................................................................................................................

14 5.3 System Pressures ...................................................................................................

15 5.4 Break .......................................................................................................................

16 5.5 Primary L oop Mass Flows .......................................................................................

17 5.6 Vessel Collapsed Liquid Levels ...............................................................................

19 5.7 Maximum Fuel Rod Surface and Core Exit Temperatures .......................................

22 5.8 Hot and Cold Legs Liquid Levels .............................................................................

23 5.9 Emergency Core Cooling Systems Mass Flow Rates ..............................................

26 5.10 Core Power .............................................................................................................

28 5.11 Voi d Fraction ...........................................................................................................

28 6 CONCLUSIONS ................................................................................................................

35 7 REFERENCES ..................................................................................................................

37

ix LIST OF FIGURES Figure 1 Schematic View of the LSTF Facility ........................................................................ 3 Figure 2 Model Nodalization ..................................................................................................

10 Figure 3 Primary and Secondary Pressures ..........................................................................

16 Figure 4 Break Mass Flow Rate ............................................................................................

17 Figure 5 Primary Loop A Mass Flow Rate. ............................................................................

18 Figure 6 Primary Loop B Mass Flow Rate. ............................................................................

18 Figure 7 Core Collapsed Liquid Level....................................................................................

19 Figure 8 Upper Plenum Collapsed Liquid Level .....................................................................

20 Figure 9 Downcomer Collapsed Liquid Level ........................................................................

21 Figure 10 Fuel Rod Surface Temperatures at Different Axial Positions. ..................................

22 Figure 11 Maximum Fuel R od Surface Temperature vs Core Exit Temperature.

.....................

23 Figure 12 Collapsed Liquid Level in the Hot Leg A. ................................................................

.24 Figure 13 Collapsed Liquid Level in the Hot Leg B24 Figure 14 Collapsed Liquid Level in the Cold Leg A.25 Figure 15 Collapsed Liquid Level in the Cold Leg B..25 Figure 16 High Pressure Injection System Mass Flow Rate.. ..................................................

26 Figure 17 Accumulator Injection System Mass Flow Rate...27 Figure 18 Low Pressure Injection System Mass Flow Rate.27 Figure 19 Core Power .............................................................................................................

28 Figure 20 Void Fraction in the LSTF at 0 s ..............................................................................

29 Figure 21 Void Fraction in the LSTF when PCT Reaches 750 K and HPI Starts30 Figure 22 Void Fraction in the LSTF when Primary Pressure = 5 MPa. ...................................

31 Figure 23 Void Fraction in the LSTF when Second PCT Excursion is Produced

.....................

32 Figure 24 Void Fraction in the LSTF at the End of the Transient33

xi LIST OF TABLES Table 1 Control Logic and Sequence of Major Events in the Experiment ................................ 6 Table 2 Core Protection Logic ................................................................................................. 7 Table 3 Vessel Nodalization. ................................................................................................... 9 Table 4 Steady-State Comparison between Experimental and Simulated Values13 Table 5 Sequence of Events Comparison between Experiment and TRACE5.. .....................14

xiii EXECUTIVE

SUMMARY

The purpose of this work is to test the capability of the thermal hydraulic code TRACE5 in the simulation of a Hot Leg Small Break LOCA (SBLOCA) in the frame of the OECD/NEA ROSA-2 Project. The main objective of this project is providing experimental data in the Large Scale Test Facility (LSTF) for assessment of thermal hydraulic computer codes. The transient considered in this work, Test 3, reproduces a PWR 1.5% hot leg SBLOCA with an assumption of total failure of High Pressure Injection (HPI) system. Test 3 was designed as counterpart test between PKL and LSTF test facilities. One of the purposes of this test is to clarify the relation between the Core Exit Temperature (CET) measured by thermocouples and the Peak Cladding Temperature (PCT) at high and low-pressure conditions corresponding to the pressure range of LSTF and PKL facilities during a hot leg SBLOCA. TRACE5 patch 2 has been used to model the LSTF and the control logic of Test 3. The model considers the high-pressure phase, the conditioning phase and the low-pressure phase performed in the LSTF experiment.

The behavior of the Pressure Vessel is analyzed, measuring the active core, upper plenum, upper head and downcomer liquid levels. Results of the simulation with TRACE5 are compared with the experimental measurements in several graphs, including primary and secondary pressures, break mass flow rate, primary mass flow rates, and collapsed liquid levels (hot leg, steam generators U-tubes, etc.).

xv ACKNOWLEDGMENTS This paper contains findings that were produced within the OECD-NEA ROSA-2 Project. The authors are grateful to the Management Board of the ROSA Project for their consent to this publication, and thank the Spanish Nuclear Regulatory Body (CSN) for the technical and financial support under the agreement STN/1388/05/748.

xvii ABBREVIATIONS AND ACRONYMS AFW Auxiliary Feedwater AIS Accumulator Injection System AM Accident Management BE Best Estimate CAMP Code Assessment and Management Program CET Core Exit Temperature CPU Central Processing Unit CRGT Control Rod Guide Tubes CSN Nuclear Safety Council, Spain DBE Design Basis Event ECCS Emergency Core Cooling System HPI High Pressure Injection IBLOCA Intermediate Break Loss-Of-Coolant Accident JAEA Japan Atomic Energy Agency JAERI Japan Atomic Energy Research Institute JC Jet Condenser LOCA Loss-Of-Coolant Accident LPI Low Pressure Injection LSTF Large Scale Test Facility MFW Main Feedwater MSIV Main Steam Isolation Valve NPP Nuclear Power Plant NRC U.S. Nuclear Regulatory Commission NV Normalized to the Steady State Value PA Auxiliary Feedwater Pump PCT Peak Cladding Temperature PF Feedwater Pump PGIT Primary Gravity Injection Tank PJ High Pressure Charging Pump PL High Pressure Injection Pump PV Pressure Vessel PWR Pressurized Water Reactor PZR Pressurizer RHR Residual Heat Removal RV Relief Valve SBLOCA Small Break Loss-Of-Coolant Accident SG Steam Generator SI Safety Injection SNAP Symbolic Nuclear Analysis Package SRV Safety Relief Valve ST Storage Tank

1 1 INTRODUCTION The purpose of this work is to describe the most relevant results achieved by using the thermal hydraulic code TRACE5 patch 2 [1 , 2] to simulate the SBLOCA transient defined in Test 3 within the OECD/NEA ROSA-2 Project (SB-HL-18 in JAEA) [3]. This transient was performed in the Large Scale Test Facility (LSTF) [4], which simulates a PWR reactor, Westinghouse type, of four loops and 3423 MW of thermal power, scaled to 1/48 in volume and two loops. Thermocouples are common ly used as an important indicator to start an Accident Management (AM) action by detecting the Core Exit Temperature (CET) excursion during reactor accidents. However, in some tests , a time delay between the detection of superheated steam by thermocouples and the CET excursion is observed [5, 6 , 7]. The CET reliability to detect core uncover during a SBLOCA is considered as one of the most important safety concerns studied in the ROSA-2 Project. Test 3 was designed as counterpart test between the PKL and LSTF test facilities. One of the purposes of this test is to clarify the relation between the CET measured by thermocouples and the fuel rod surface temperature (PCT) at high and low-pressure conditions corresponding to the pressure range of LSTF and PKL facilities during a hot leg SBLOCA. Another goal of this counterpart test is to study two large integral test facilities with different designs to draw technical findings useful for solving scaling problems.

3 2 LSTF FACILITY DESCRIPTION In this section, a brief description of the LSTF facility [10] is presented. LSTF simulates a PWR reactor, Westinghouse type, of four loops and 3423 MW of thermal power. The facility is electrically heated, scaled 1:1 in height and 1:48 in flow areas and volumes, with exception of the loops, which are defined by a scaling factor of 1:24 in flow areas and volumes. Figure 1 shows the scheme of the LSTF facility. As it can be seen, the primary coolant system consists of the Pressure Vessel (PV) and two symmetrical primary loops: loop A with the pressurizer (PZR) and loop B. Figure 1 Schematic View of the LSTF Facility Each loop contains a primary Coolant Pump (PC) and a Steam Generator (SG). The secondary-coolant system consists of the jet condenser (JC), the Feedwater Pump (PF), the Auxiliary Feedwater Pumps (PA) and related piping system in addition to two SG secondary systems. The Pressure Vessel (PV) is composed of an upper head above the upper core support plate, the upper plenum between the upper core support plate and the upper core plate, the core, the lower plenum and the downcomer annulus region surrounding the core and upper plenum. LSTF vessel has 8 upper head spray nozzles (of 3.4 mm inner-diameter), and 8 Control Rod Guide Tubes (CRGTs) characterize the flow path between the upper head and upper plenum. The maximum core power of the LSTF is limited to 10 MW which corresponds to 14% of the volumetrically scaled PWR core power and is sufficiently capable to simulate PWR decay heat power after the reactor scram. Regarding the SGs, each of them contains 141 U-tubes which can be classified into separate groups depending on their length. The U-tubes have an inner diameter of 19.6 mm and an outer diameter of 25.4 mm (with 2.9 mm wall thickness). On the other hand, vessel, plenum and riser of steam generators have a height of 19.840, 1.183 and 17.827 m, respectively. The downcomer is 14.101 m in height.

5 3 TRANSIENT DESCRIPTION Test 3 reproduces a PWR hot leg Small Break LOCA, which flow area corresponds to 1.5% of reference PWR cold leg area, with High Pressure Injection (HPI) into the PV upper plenum [3].

The transient performed in LST F is divided in to three phases; high-pressure transient phase, conditioning phase, and low-pressure transient phase to meet the PKL pressure. The complete control logic of the transient is listed in Table 1. The transient starts at time 0 with opening the break valve in the hot leg of loop without pressurizer and increasing the rotational speed of the coolant pumps. Few seconds afterwards, the scram signal is generated. This signal produces the initiation of the core power decay curve. In addition, the scram signal produces the initiation of the primary coolant pumps coast, the turbine trip, the closure of the Main Steam Isolation Valves (MSIV) and the termination of the Main Feedwater (MFW).

To protect the facility, the LSTF Core Protection System automatically decreases the core power when the maximum fuel rod surface temperature reaches 958 K, as it can be seen in Table 2. Immediately after the maximum fuel rod surface temperature reaches 750 K, the HPI system injects coolant into the PV upper plenum to avoid subcooled water layer being formed at the PV bottom. This phase is terminated when the primary pressure decreases to 5 MPa and the break valve is temporarily closed. In the conditioning phase, the core power is manually changed to a constant value (1.16 MW). The primary mass inventory is recovered by the continuous HPI injection into the PV upper plenum. When the hot leg liquid level recovers up to the middle level, the HPI is terminated. The Relief Valves (RVs) are fully opened in both SGs for depressurization. The Auxiliary Feedwater (AFW) is then injected in to both SGs to avoid significant liquid level drop. When the primary pressure decreases to 3.9 MPa, the RVs are closed and AFW is terminated in both SGs. This phase is finished when the primary pressure reaches 4.5 MPa. In the low-pressure phase, the break valve is again opened. Due to the coolant loss through the break, the core uncover is produced. Immediately after the CET reaches 623 K, the SG secondary-side depressurization is initiated by fully opening the RVs at both SG as an AM action. The AFW is also injected in to both SGs. The Accumulator Injection system (AIS) is initiated when the primary pressure reaches 2.6 MPa.

The Low Pressure Injection (LPI) system is actuated when the PV lower plenum pressure reaches 1 MPa. The transient is terminated when continuous core cooling by the LPI system is confirmed.

6 Table 1 Control Logic and Sequence of Major Events in the Experiment Break. Time zero High-pressure transient phase Generation of scram signal.

Primary pressure = 12.97 MPa Pressurizer (PZR) heater off.

Generation of scram signal or PZR liquid level below 2.3 m Initiation of core power decay curve.

Generation of scram signal.

Initiation of Primary Coolant Pump coastdown. Generation of scram signal.

Turbine trip (closure of steam generators Main Steam Isolation Valve s). Generation of scram signal.

Closure of steam generators Main Steam Isolation Valves.

Generation of scram signal.

Termination of steam generators Main Feedwater.

Generation of scram signal.

Generation of Safety Injection (SI) signal. Primary pressure = 12.27 MPa Initiation of High Pressure Injection system (HPI) into the pressure vessel upper plenum.

Maximum fuel rod surface temperature = 750 K

. Initiation of steam generators secondary-side depressurization by fully opening of Relief Valves in both loops as AM action.

Maximum core exit temperature = 623 K Low-pressure transient phase Initiation of Auxiliary Feed Water in both loops Initiation of AM action.

Initiation of Accumulator Injection system in both loops Primary pressure = 2.6 MPa Termination of Accumulator Injection system in both loops Primary pressure =1.2 MPa Initiation of Low Pressure Injection system in both loops PV lower plenum pressure = 1 MPa 7 Table 2 Core Protection Logic Core power to Maximum fuel rod surface temperature (K) 70% 958 K 35% 961 K 13% 966 K 5% 977 K 0% (core power trip) 1003 K

9 4 TRACE5 MODEL OF LSTF LSTF has been modeled with 97 hydraulic components (11 BREAKs, 12 FILLs, 25 PIPEs, 2 PUMPs, 1 PRIZER, 26 TEEs, 19 VALVEs and 1 VESSEL). Figure 2 shows the nodalization of the model using the Symbolic Nuclear Analysis Package software (SNAP) [9]. Primary side comprises cold and hot legs, pumps, loop seals, a pressurizer in loop A, the ECCS which includes AIS, HPI and LPI systems, the U-tubes of both SG and the PV. On the other hand, secondary side includes steam separators, risers, downcomers, Safety Relief Valves , Main Steam Isolation Valves and FILLs to provide Main Feedwater and Auxiliary Feedwater. The pressure vessel has been modelled using a 3DVESSEL component. The VESSEL nodalization consists of 20 axial levels, 4 radial rings and 10 azimuthal sectors. For each axial level, volume and effective flow area fractions have been set according to technical specifications provided by the organization [4]. Table 3 shows the vessel nodalization in the axial direction. Table 3 Vessel Nodalization Levels Parts of the vessel 1-2Lower plenum 3-11Core 12 Upper core plate 13-16Upper plenum 17 Upper core support plate 18-20Upper head The 3D-VESSEL is connected to different 1D components: 8 Control Rod Guide Tubes (CRGT), hot leg A and B (level 15), cold leg A and B (level 15) and a bypass channel (level 14). The CRGTs have been simulated by PIPE components, connecting levels 14 and 19 and allowing the flow between upper head and upper plenum. 30 HTSTR components simulate the fuel assemblies in the active core. A POWER component manages the power supplied by each HTSTR to the 3D-VESSEL. Fuel elements (1008 in total) were distributed into the 3 rings: 154 elements in ring 1, 356 in ring 2 and 498 in ring 3. In both axial and radial direction, different peaking factors have been considered. The power ratio in the axial direction presents a peaking factor of 1.495, while the radial power profile is divided into three power zones using the first three radial rings. Depending on the radial ring, different peaking factors have been considered (0.66 in ring 1, 1.51 in ring 2 and 1.0 in ring 3). The number of fuel rod components associated with each heat structure has been determined from the technical documentation given, considering the distribution of fuel rod elements in the vessel.

10 A detailed model of SG (geometry and thermal features) has been developed. Boiler and downcomer components of secondary-side have been modeled by TEE components. U-tubes have been classified into three groups according to each average length. Figure 2 Model Nodalization The steam separator can be simulated in TRACE5 setting a friction coefficient (FRIC) greater than 10 22 at a determined cell edge, allowing only gas phase to flow through the cell interface. Heat transfer between primary and secondary sides has been performed using HTSTR components. Cylindrical-shape geometry has been used to best fit heat transmission. Inner and outer surface boundary conditions for each axial level have been set to couple HTSTR component to hydro components (primary and secondary fluids). Different models varying the number of U-tube groups were tested (1, 3 and 6 groups). It was found that the results do not apparently change, using these models. Heat losses to the environment have been considered in the secondary-side walls.

U-tubesLoop B U-tube sPressurizer Accumulators LPI LPI Secondary side A Secondary side B Loop A Break Valve HPI during Test 3 H PI H PI 11 Choke model predicts for a given cell the conditions for which choked flow is expected to occur, providing three different models: subcooled-liquid, two-phase and single-phase vapor model. TRACE5 patch 2 code allows to choose the subcooled-liquid and two-phase coefficients. In this case, the default values (1.0) have been selected. The break has been simulated by means of a VALVE component connected to a BREAK component to establish the boundary conditions.

13 5 RESULTS AND DISCUSSION 5.1 Steady- tate Steady-state conditions achieved in the simulation w ere in reasonable agreement with the experimental values. It can be seen i n Tabl e 4, where the relative error s (%) between experimental and simulated result s for different item s are listed.

To achieve the steady state conditions, the duration of simulation was stated t o 1000 s. Tabl e 4 Steady-St ate Condition Comparison between Experimental an d Simulated Values Item (Loop with PZR)

Relative Error (%) Core Power 0.00 Hot leg Fluid Temperature 0.10 Cold leg Fluid Temperature 0.27 Mass Flow Rate 3.68 Pressurizer Pressure 0.06 Pressurizer Liquid Level 3.60 Accumulator System Pressure

-0.38Accumulator System Temperature

-0.56SG Secondary

-side Pressure 0.96 SG Secondary

-side Liquid Level 5.67 Steam Flow Rate 3.86 Main Feedwater Flow Rate 3.40 Main Feedwater Temperature 0.06 Auxiliary Feedwater Temperature

-0.13 14 5.2 Transient Tabl e 5 list s the chronological sequence of the transient event s and the comparison between the experiment and TRACE. Tabl e 5 Chronological Sequence of Events Comparison between Experiment and TRACE5 Event Experiment Time (s) TRACE5 Time (s) Break valve open 0 0 Scram signal 29 32 Closure of steam generators (SG) Main Steam Isolation Valves 32 33 Initiation of coastdown of primary coolant pumps 33 35 Termination of SG main feedwater 34 33 SI signal 37 43 Initiation of core power decay 50 47 Primary coolant pumps stop 281 280 Primary pressure lower than secondary side pressure 1310 1317 T he increasing in fuel rod surface temperature starts 1595 1680 Maximum fuel rod surface temperature reached 750 K 1840 1796 Initiation of High Pressure Injection (HPI) into the PV upper plenum 1844 1796 Break valve closure 2172 2113 Manual change of core power to a constant value 2215 2500 Termination of HPI system into the PV upper plenum 2852 2880 Initiation of SG secondary

-side depressurization by fully opening Relief Valves (RVs) in both loops 2880 2893 Initiation of Auxiliary Feed Water (AFW) in both loops 2900 2893 Termination of SG secondary

-side depressurization 3028 2959 Termination of AFW in both loops 3055 2959 Break valve open again 3323 3303 Start of the increasing in fuel rod surface temperature 3983 4203 15 Primary pressure becomes lower than SG secondary side pressure 4105 4250 Maximum core exit temperature = 623 K 4390 4507 Initiation of SG secondary

-side depressurization by fully opening RVs in both loops as AM action 4394 4519 Initiation of AFW in both loops 4410 4519 Maximum fuel rod surface temperature 4413 4545 Initiation of Accumulator Injection system (AIS) in both loops 4500 4590 Termination of AIS in both loops 4829 4786 Initiation of Low Pressure Injection (LPI) system in both loops 5003 4830 5.3 System Pressures A comparison between primary an d secondary pressures i s presented i n Figure 3. In the high-pressure phase (until 2100 s), the primary pressure start s t o decrease when the brea k valve is opened due t o the coolant discharged through the break. The scram signal i s generated when the primary pressure i s low er than 12.97 MPa. The SI signal is generated when the primary pressure reache s 12.27 MPa. The generation of the scr am signal cau ses the closure of SG MSIVs an d the beginning of the primary coolant pumps coastdow

n. The SG secondary-side pressur e rapidly increases after the closure of MSIVs. From thi s moment on, the SG secondary-side pressure start s to oscill ate by opening and closing the RV s of both SGs. The primary pressure becomes low er tha n the SG seconda ry-side pressure at about 1 250 s soon after the brea k flow turns into single-phase steam (Figure 4) and decreases until 5 MPa (at 2170 s), when the brea k valve is closed.

I n the conditioning phase (until 33 00 s), the primary pressure increases once up to 6.5 MP a and decreases to 4 MPa following the SG secondary-side depressurizatio

n. The primary and secondary pressures increase again up to about 4.5 MPa after the termination of the SG secondaryside depressurization. The low-pressure pha se start s by opening the break valve again (at 3300 s). The primary pressure becomes low er than the SG pressure at about 4100 s, slightly aft er the core boil-off is produced. Immediately after the CET reaches 623 K (Figure 10), the SG secondaryside depressurization is initiated. Then, the primary pressure decreases following the SG pressure and activates the Accumulator Injection and the Low Pressur e Injection systems.

I n general, both primary and secondary-side pres sures ar e successfully reproduced by TRACE5 in the whole transient. However, some discrepanci es are observed regarding the pressure drop slopes, which ar e different in all the cases.

16 Figure 3 Primary and Secondary Pressures 5.4 Break Figure 4 show s the mass flow rate throug h the bre ak. I n the high-pressure phase, the brea k flow rate decrease s when the brea k flow turns from single-phase liqui d t o two-phase flow (200 s). At 1250 s, the brea k flow tur ns t o single-phase vapor. To adjust the break ma ss flow rate with TRACE 5, a sensitivity analysi s varying the discharge coefficient of the Choked Flow model was performed. I n the results show n, the discharge coefficients have been fixed to 1.0. During the interval between 2 50 and 1 000 s, the mass flow rate through the break obtained with TRACE5 is higher than the experimental values. However, the changes from liquid to two-phase and fr om two-phase to one phase vapor are reproduced at similar time. LOW PRESSURE PHASE CONDITIONING PHASE Scram signal MSIVs closure RV cycling open/close Break valve closure 1 st depressurization 2 nd depressurization HPI LPI starts HIGH PRESSURE PHASE AIS starts

17 Figure 4 Break Ma ss Flow Rate 5.5 Primary Loop Ma ss Flows Figures 5 and 6 show the primary mass flow rate in both loops. During the first second s of the transient, the primary ma ss flow increases du e t o the higher angul ar speed of the pumps. Then, the primary mass flow rate in both loops decreases according to the pump coastdown. The main discrepanci es are found during the accumulator water injection. At 4500 s, the experimental primary ma ss flow rate in both loops are higher tha n the simulation results du e to the accumulator water entrance i n the cold legs. Thi s effect ha s not bee n properly predicted by TRACE5. LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

18 Figure 5 Primary Loop A M ass Flow Rate Figure 6 Primary Loop B M ass Flow Rate LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

19 5.6 Vessel Collapsed Liquid Levels The Figur es 7 , 8 and 9 show a comparison between the collapsed liquid level s in the core, upper plenum an d downcomer, respectively, obtaine d for both experimental an d simulation results. I n the core liquid level, a significant drop start s due to the boil-off at about 1550 s, when the upper plenum i s emptied. The core uncovering takes plac e after the primary pressure becomes low er than the secondary pressure. The collapsed liquid level continues to drop to about 1/3 of the active core len gth until 1870 s, even after the initiation of the high-pressure coolant injection i nto the PV upper plenum.

I n the conditioning phase, the liquid level in the core and the upper plenum i s recov ered du e to the high-pressure coolant injection. When the upp er plenum reache s the middle level (at 2850 s), the HPI finishes. Figure 7 Core Collapsed Liquid Level When the break valve is opened again at 3300s to start the low-pressure phase, the liquid level drops in the upper plenum and the core. The core liquid level begins to drop at about 3900 s, and the core uncovering is produced before the primary pressure becomes lower than the secondary pressure. The core liquid level is recovered following the primary depressurization. LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

20 Figure 8 Upper Plenum Collapsed Liquid Level Regarding the downcomer liquid level in the high pressure phase, it drops gradually up to 1800 s. It is recover ed at 1910 s, after the initiation of the HPI in to the PV upper plenum. In the conditioning phase, the temporary liquid level drop happens as in the upper plenum and the core due to the secondary depressurization. In the low-pressure phase (after 35 00 s), the liquid level drops following the core boil-off. The downcomer liquid level is steeply recover ed at 4530 s due to the accumulator coolant injection. As it can be observed in Figure 9 , some discrepancies appear during the first emptying of the downcomer, which is delayed in the simulation about 500 s. LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

21 Figure 9 Downcomer Collapsed Liquid Level LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

22 5.7 Maximum Fuel Ro d Surface and Core Exit Temperatures Figure 10 show s a comparison between nine axial position s of a fuel rod. A HTSTR from the second radial ring has been chosen. The axial position s correspond to the nine axial level s in which is divided the core. The main discrepancies are observed during the second temperature excursion. TRACE5 doe s not reproduce the maximum value, obtaining a pe ak 100 K lower. In the high-pressure phase, the PCT reache s 750 K at 1840 s and the high pressure injection into the PV upper plenum is initiated. Figure 10 Fuel R od Surface Temperatures at D ifferent Axial Positions In the experiments performed by the NEA Working Group on the Analysis and Management of Accidents [7], a significant difference between CET and PCT was observed in the measurements obtained in all the facilities. Thus, as the CET triggers the AM actions, but the safety variable normally followed in nuclear safety is the PCT, a more detailed study of both variables and the relation between them was suggested. Figure 11 shows the representation CET versus PCT. This figure allows clarifying the relation between CET measured by thermocouples and PCT during a hot leg SBLOCA. This relation allows obtaining the PCT that corresponds to a determined CET. The delay between the PCT and the CET excursions produces that the CET remains constant while the PCT increases to 600 K. However, in TRACE5 both excursions are initiated at the same time. LOW PRESSURE PHAS E CONDITIONING PHASE HIGH PRESSURE PHASE

23 Figure 11 Maximum Fuel Rod Surface Temperature vs Core Exit Temperature 5.8 Hot and Cold Legs Liquid Levels Figures 12 and 13 show the liquid level i n bot h hot legs. A s it can be seen, the liquid level is almost the same in bot h legs (intact and broken loop). In the high-pressure phase, the hot leg fluid becomes saturated at 53 s. The liquid level was kept at around 3/4 to 1/2 of the inner diameter until about 1310 s , when the steam bre ak flow and primary pres sure decreases. The hot leg becomes empty at about 1400 s, when the liquid level in the upper plen um begins to drop. In the conditioning phase, the hot leg liquid level is recover ed at abo ut 2640 s after the core reflooding and reache s the middle level. In the low-pressure phase, the liquid level start s to decrea se just after the brea k valve open. The hot leg becomes empty at abo ut 3600 s. The hot leg liquid level i s recover ed at 4530 s du e to the accumulator injection.

In high-and low-pressure phases, the fluid is kept saturated i n the intact loop and superheated i n the broken loop during the core uncovering. It suggest s that t he steam preferentially flow s tow ards the brea k from the PV upper plenum while stagnat es i n the loo p A, except during the SG depressurization. Figures 14 and 15 show the cold leg liquid levels.

In loop A, the cold leg becomes empty at about 900 s, wherea s in loop B it occur s at 1250 s. However, in the simulation bot h cold legs are empty at the sam e tim e (around 1250 s). During the coolant injection by the accumulator and LPI systems, the liquid level i s recov ered in both loops. However, TRACE does not exactly reproduce the recovering of the liquid level in both cold legs, due to the rapid coolant discharge throug h the break from the accumulation injection.

24 Figure 12 Collapsed Liquid Level in the Hot Leg A Figure 13 Collapsed Liquid Level in the Hot Leg B LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

25 Figure 14 Collapsed Liquid Level in the Cold Leg A Figure 15 Collapsed Liquid Level in the Cold Leg B LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

26 5.9 Emergency Core Cooling Systems Ma ss Flow Rate The HPI into the PV upp er plenum i s activated immediately after the maximum fuel rod surface temperatur e reaches 750 K (Figur e 10). Thi s HPI configuration i s use d t o avoid thermal stratification occurring i n the PV low er plenum. The coolant injection finishes at 2850 s, when the liquid level in the hot leg recov ers the middle level. A s it can be seen in Figure 16, wher e the HPI ma ss flow rates ar e show n, TRACE5 reproduces the HPI injection successfully. Figure 16 High Pressure Injection System Mass Flow Rate Figure 17 shows the coolant injection flow rate from the Accumulator Injection System. As it can be seen, the simulated mass flow rate is lower than the experimental data, due to the different primary pressure drop. This fact could explain the differences in hot and cold legs refill. The LPI system is activated in both loops at 5000 s. In the simulation, this event is advanced as it can be seen in Figure 18. LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

27 Figure 17 Accumulator Injection System Mass Flow Rate Figure 18 Low Pressure Injection System Mass Flow Rate LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE

28 5.10 Core Power Figure 19 shows the experimental and the simulated core power curves. The core power starts to decay at 50 s following the core power curve decay. The core power is manually changed to a constant value of 1.16 MW at 2215 s in the experiment and at 2500 s in the simulation. As it can be seen, the simulated core power curve has a good agreement with the experimental curve. Figure 19 Core Power 5.11 Void Fraction Figures 20, 21, 22, 23 and 24 show the void fraction achieved using the LSTF TRACE5 model during the transient, when important events happen. Figure 19 shows the void fraction in the LSTF at the initiation of the test. As it can be seen, primary and secondary sides are full of liquid at this time. Figure 20 shows the void fraction when the PCT reaches 750 K and the HPI injection starts. At this time, the pressurizer is empty, and the liquid is located in the loop seals, accumulators, bottom of the PV and SGs. Figure 21 shows the void fraction when the primary pressure drops to 5 MPa and the break valve is closed. In this moment, the lower plenum of the PV, SG and AIS remain full of liquid, while the loop seals are almost empty. The situation when the second maximum of the PCT is reached is shown in Figure 22. SGs are empty while the loop seals LOW PRESSURE PHASE CONDITIONING PHASE HIGH PRESSURE PHASE Power (W) 29 have some liquid. Finally, Figure 23 shows the void fraction at the end of the transient. As it can be seen, the SGs are almost empty and the accumulators have not been completely emptied. Figure 20 Void Fraction in the LSTF at 0 s 30 Figure 21 Void Fraction in the LSTF when PCT Reaches 750 K and HPI Starts 31 Figure 22 Void Fraction in the LSTF when Primary Pressure = 5 MPa 32 Figure 23 Void Fraction in the LSTF when Second PCT Excursion is Produced 33 Figure 24 Void Fraction in the LSTF at the End of the Transient

35 6 CONCLUSIONS Results show that TRACE5 can successfully reproduce all the phenomena produced in Test 3 OECD/NEA ROSA-2 Project (SB-HL-18 in JAEA) during the different transient phases: high-pressure, conditioning, and low-pressure. The main variables of the system present a good agreement in comparison to experimental data. System pressures, collapsed liquid levels in the PV , CET and PCT excursions are well reproduced. However, some discrepancies observed in the break mass flow rate could be attributed to the lack of a single-phase vapor coefficient in the choked flow model of TRACE5 patch 2. These differences can affect the mass flow rate through the hot and cold legs. Some discrepancies are also found in the maximum values reached during the second temperature excursions. However, these discrepancies do not affect the relation between the core exit temperature and the fuel rod surface temperature.

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