ML19066A174

NUREG/IA-0502, International Agreement Report - Post-Test Analysis of Cold Leg Small Break 4.1% at PSB-VVER Facility Using Trace V5.0

ML19066A174
Person / Time
Issue date:	03/31/2019
From:	Heralecky P, Kirk Tien Office of Nuclear Regulatory Research, TES, Ltd
To:
Meyd, Donald
References
NUREG/IA-0502
Download: ML19066A174 (89)

Text

NUREG/IA-0502 Post-Test Analysis of Cold Leg Small Break 4.1% at PSB-VVER Facility using TRACE V5.0 Prepared by:

Petr Heralecky TES Ltd.

Prazska 597 674 01 Trebic, Czech Republic K. Tien, NRC Project Manager Division of Systems Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Manuscript Completed: April 2018 Date Published: March 2019 Prepared as part of The Agreement on Research Participation and Technical Exchange Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by U.S. Nuclear Regulatory Commission

AVAILABILITY OF REFERENCE MATERIALS IN NRC PUBLICATIONS NRC Reference Material Non-NRC Reference Material As of November 1999, you may electronically access Documents available from public and special technical NUREG-series publications and other NRC records at libraries include all open literature items, such as books, NRCs Library at www.nrc.gov/reading-rm.html. Publicly journal articles, transactions, Federal Register notices, released records include, to name a few, NUREG-series Federal and State legislation, and congressional reports.

publications; Federal Register notices; applicant, Such documents as theses, dissertations, foreign reports licensee, and vendor documents and correspondence; and translations, and non-NRC conference proceedings NRC correspondence and internal memoranda; bulletins may be purchased from their sponsoring organization.

and information notices; inspection and investigative reports; licensee event reports; and Commission papers Copies of industry codes and standards used in a and their attachments. substantive manner in the NRC regulatory process are maintained at NRC publications in the NUREG series, NRC regulations, The NRC Technical Library and Title 10, Energy, in the Code of Federal Regulations Two White Flint North may also be purchased from one of these two sources. 11545 Rockville Pike Rockville, MD 20852-2738

1. The Superintendent of Documents U.S. Government Publishing Office These standards are available in the library for reference Mail Stop IDCC use by the public. Codes and standards are usually Washington, DC 20402-0001 copyrighted and may be purchased from the originating Internet: bookstore.gpo.gov organization or, if they are American National Standards, Telephone: (202) 512-1800 from Fax: (202) 512-2104 American National Standards Institute 11 West 42nd Street

2. The National Technical Information Service New York, NY 10036-8002 5301 Shawnee Road www.ansi.org Alexandria, VA 22312-0002 (212) 642-4900 www.ntis.gov 1-800-553-6847 or, locally, (703) 605-6000 Legally binding regulatory requirements are stated only in A single copy of each NRC draft report for comment is laws; NRC regulations; licenses, including technical available free, to the extent of supply, upon written specifications; or orders, not in NUREG-series publications.

request as follows: The views expressed in contractor prepared publications in this series are not necessarily those of the NRC.

Address: U.S. Nuclear Regulatory Commission The NUREG series comprises (1) technical and adminis-Office of Administration trative reports and books prepared by the staff (NUREG-XXXX) or agency contractors (NUREG/CR-XXXX), (2)

Multimedia, Graphics, and Storage &

proceedings of conferences (NUREG/CP-XXXX), (3) reports Distribution Branch resulting from international agreements (NUREG/IA-XXXX),

Washington, DC 20555-0001 (4) brochures (NUREG/BR-XXXX), and (5) compilations of E-mail: distribution.resource@nrc.gov legal decisions and orders of the Commission and Atomic Facsimile: (301) 415-2289 and Safety Licensing Boards and of Directors decisions under Section 2.206 of NRCs regulations (NUREG-0750).

Some publications in the NUREG series that are posted at NRCs Web site address www.nrc.gov/reading-rm/ DISCLAIMER: This report was prepared under an doc-collections/nuregs are updated periodically and may international cooperative agreement for the exchange of technical information. Neither the U.S. Government nor any differ from the last printed version. Although references to agency thereof, nor any employee, makes any warranty, material found on a Web site bear the date the material expressed or implied, or assumes any legal liability or was accessed, the material available on the date cited responsibility for any third partys use, or the results of such may subsequently be removed from the site. use, of any information, apparatus, product or process disclosed in this publication, or represents that its use by such third party would not infringe privately owned rights.

NUREG/IA-0502 Post-Test Analysis of Cold Leg Small Break 4.1% at PSB-VVER Facility using TRACE V5.0 Prepared by:

Petr Heralecky TES Ltd.

Prazska 597 674 01 Trebic, Czech Republic K. Tien, NRC Project Manager Division of Systems Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Manuscript Completed: April 2018 Date Published: March 2019 Prepared as part of The Agreement on Research Participation and Technical Exchange Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by U.S. Nuclear Regulatory Commission

ABSTRACT The U.S. NRC best estimate thermo-hydraulic computer code TRACE V5.0 has been assessed against the 4.1 % cold leg break (CL-4.1-03) experiment at the large-scale test facility PSB-VVER. The PSB-VVER facility is a 1:300 volume-power scaled model of VVER-1000 NPP located in Electrogorsk, Russia. An extensive TRACE input deck of PSB-VVER facility was developed. The model includes all important components of the PSB-VVER facility: reactor, 4 separated loops, pressurizer, HPIS and LPIS ECCS, several break units, main circulation pumps, steam generators, and important parts of secondary circuit. The TRACE (TRAC/

RELAP Advanced Computational Engine) is the latest in a series of advanced, best-estimate reactor systems codes developed by the U.S. Nuclear Regulatory Commission in frame of CAMP (Code Application and Maintenance Program). The TRACE code is a component-oriented reactor systems analysis codes designed to analyze light water reactor transients up to the point of significant fuel damage. The original validation of the TRACE code was mainly based on experiments performed on experimental facilities of typical PWR design. There are some different features of VVER design comparing to PWR. Therefore the validation of the thermo-hydraulic codes for VVER types of reactors is often required by national regulators.

The presented analysis is the latest in series of TRACE and RELAP5 assessment calculations evaluated at the company TES. The purpose of performed analyses is to extend the validation of the TRACE code focused on VVER type of NPPs and to support applicability of the TRACE code in the Czech Republic.

iii

TABLE OF CONTENTS ABSTRACT ................................................................................................................... iii LIST OF FIGURES........................................................................................................ vii LIST OF TABLES .......................................................................................................... ix EXECUTIVE

SUMMARY

............................................................................................... xi ACKNOWLEDGMENTS .............................................................................................. xiii ABBREVIATIONS AND ACRONYMS.......................................................................... xv 1 INTRODUCTION .................................................................................................... 1-1 2 FACILITY AND TEST DESCRIPTIONS ................................................................. 2-1 2.1 PSB-VVER Facility ...................................................................................................... 2-1 2.2 Experiment CL-4.1-03 ................................................................................................. 2-4 2.2.1 Facility configuration ........................................................................................ 2-4 2.2.2 Initial Conditions .............................................................................................. 2-4 2.2.3 Boundary Conditions (test scenario) ............................................................... 2-5 3 THE TRACE V5.0 CODE ........................................................................................ 3-1 3.1 The TRACE V5.0 code assessment ........................................................................... 3-2 3.2 VVER typical features related to TRACE V5.0 code assessment ............................... 3-2 4 INPUT DECK DESCRIPTION................................................................................. 4-1 4.1 The TRACE input deck ............................................................................................... 4-1 5 RESULTS.... ........................................................................................................... 5-1 5.1 Steady-state calculation .............................................................................................. 5-1 5.2 Transient calculation ................................................................................................... 5-3 5.2.1 Time course of the transient ............................................................................... 5-3 5.2.2 Comparison of TRACE prediction to reference data .......................................... 5-5 5.2.3 Sensitivity studies performed ............................................................................. 5-6 5.3 Quantitative Assessment of the Calculations .............................................................. 5-7 6 RUN STATISTICS .................................................................................................. 6-1 7 CONCLUSIONS ..................................................................................................... 7-1 8 REFERENCES ....................................................................................................... 8-1 APPENDIX A: INPUT DECK NODALISATION SCHEMES ...................................... A-1 APPENDIX B: APPENDIX B THERMAL-HYDRAULIC DIAGRAM AND MEASUREMENT LOCALISATION AT PSB-VVER FACILITY ......... B-1 APPENDIX C: COMPLETE SET OF COMPARISON PLOTS FOR TRACE CALCULATION ................................................................................. C-1 v

LIST OF FIGURES Figure 1 General View of PSB-VVER Facility ................................................................. 2-2 Figure A-1 TRACE Nodalization Scheme of Primary Circuit of PSB-VVER Facility .......... A-2 Figure A-2 TRACE Nodalization Scheme of Secondary Circuit of PSB-VVER Facility ..... A-3 Figure B-1 PSB-VVER Thermal-Hydraulic Diagram .......................................................... B-2 Figure B-2 PSB-VVER Reactor Model Measurements...................................................... B-3 Figure B-3 PSB-VVER Loop 1 and SG-1 Model Measurement ......................................... B-4 Figure C-1 Primary Pressure ............................................................................................. C-2 Figure C-2 Fuel Cladding Temperature (Top Of The Core) ............................................... C-2 Figure C-3 Pressurizer Level ............................................................................................. C-3 Figure C-4 LPIS Flow (Boundary Condition)...................................................................... C-3 Figure C-5 Accumulators Levels ........................................................................................ C-4 Figure C-6 Accumulators Pressures .................................................................................. C-4 Figure C-7 Fuel Rod Simulator Power (Boundary Condition) ............................................ C-5 Figure C-8 Core By-Pass Power (Boundary Condition)..................................................... C-5 Figure C-9 Secondary Side Pressures .............................................................................. C-6 Figure C-10 Steam Generators Levels ................................................................................ C-7 Figure C-11 Loop 1 Temperatures....................................................................................... C-8 Figure C-12 Loop 2 Temperatures....................................................................................... C-9 Figure C-13 Loop 3 Temperatures..................................................................................... C-10 Figure C-14 Loop 4 Temperatures..................................................................................... C-11 Figure C-15 Loop 1 Pressure Differences 1 ...................................................................... C-12 Figure C-16 Loop 1 Pressure Differences 2 ...................................................................... C-13 Figure C-17 Loop 2 Pressure Differences 1 ...................................................................... C-14 Figure C-18 Loop 2 Pressure Differences 2 ...................................................................... C-15 Figure C-19 Loop 3 Pressure Differences 1 ...................................................................... C-16 Figure C-20 Loop 3 Pressure Differences 2 ...................................................................... C-17 Figure C-21 Loop 4 Pressure Differences 1 ...................................................................... C-18 Figure C-22 Loop 4 Pressure Differences 2 ...................................................................... C-19 Figure C-23 Pressure Differences DP01-DP04 (DC)......................................................... C-20 Figure C-24 Pressure Differences DP05-DP08 (Lower Plenum + Lower Part of FRS) ..... C-21 Figure C-25 Pressure Differences DP09-DP12 (FRS + Lower Part of Upper Plenum) ..... C-22 Figure C-26 Pressure Differences DP13-DP16 (Upper Part of Upper Plenum) ................ C-23 Figure C-27 Primary Loops Flow ....................................................................................... C-24 Figure C-28 Break Flow ..................................................................................................... C-25 Figure C-29 Reactor Collapsed Level ................................................................................ C-25 vii

LIST OF TABLES Table 1 Main Operational Characteristics of PSB-VVER Comparing to VVER-1000 .......... 2-3 Table 2 Test Facility Configuration in CL-4.1-03 Test.......................................................... 2-4 Table 3 Measured Initial Condition for CL-4.1-03 Test ........................................................ 2-4 Table 4 Main Events During CL-4.1-03 Test........................................................................ 2-5 Table 5 Time Dependence of Dimensionless Core Power .................................................. 2-6 Table 6 TRACE Components Statistic ................................................................................. 4-2 Table 7 List of the Main Systems and Components of PSB-VVER TRACE Input Deck ...... 4-3 Table 8 Initial Conditions (TRACE Calculation vs. Experiment Comparison) ...................... 5-1 Table 9 Chronology of main events (TRACE calculation vs experiment comparison) ......... 5-5 Table 10 ACAP metrics settings ............................................................................................ 5-7 Table 11 Acceptability criterions ............................................................................................ 5-7 Table 12 Evaluation of FOM for CL-4.1-03 test .................................................................... 5-9 Table 13 Run statistics .......................................................................................................... 6-1 Table 14 Final Average FOM Evaluated for Pre-defined Time Windows...............................7-1 ix

EXECUTIVE

SUMMARY

The purpose of this work is to contribute to the USNRC thermal-hydraulic codes assessment as agreed in the CAMP agreement. TRACE V5.0 Patch 3 and Patch 4 computer codes are assessed against LOCA test data from the PSB-VVER test facility.

xi

ACKNOWLEDGMENTS This report and used input model of the Integral Test Facility (ITF) PSB-VVER is based on the outcomes of the project FI-IM5/150 that was funded by The Czech Ministry of Industry and Trade in frame of national RD activities.

xiii

ABBREVIATIONS AND ACRONYMS ACC Hydro-accumulator ACAP Automated Code Assessment Program ADS Atmospheric Dump System ATWS Anticipated Transient Without SCRAM CAMP Code Application and Maintenance Program CCFL Counter Current Flow Limitation CFD Computational Fluid Dynamics CL Cold Leg CPU Central Processing Unit CS Core Simulator DC Downcomer ECCS Emergency Core Cooling System ECI Exterior Communications Interface EREC Electrogorsk Research and Engineering Institute FOM Figure of Merit FRS Fuel Rod Simulator GNU GNU's Not Unix!

HL Hot Leg HPIS High Pressure Injection System IE Initiating Event ITF Integral Effect Test Facility LOCA Loss of Coolant Accident LP Lower Plenum LPIS Low Pressure Injection System LWR Light Water Reactor MCP Main Coolant Pump NEA Nuclear Energy Agency NPP Nuclear Power Plant NRC Nuclear Regulatory Commission OECD Organization for Economic Co-operation and Development PARCS Purdue Advanced Reactor Core Simulator PKL Experimental Facility in Finland PCS Plant Control System PRZ Pressurizer PRISE Primary to Secondary Leak PWR Pressurized Water Reactor RD Research and Development RPV Reactor Pressure Vessel SCRAM Emergency Shutdown of a Nuclear Reactor SETF Separate Effect Test Facility SETS Semi-implicit Method and the Stability Enhancing Two-Step xv

SG Steam Generator SNAP Symbolic Nuclear Analysis Package TC Thermocouple TH Thermal Hydraulic UP Upper Plenum VVER Russian Pressurized Water Type Reactor xvi

1 INTRODUCTION The assessment of LWR best estimate thermo-hydraulic computer codes is mainly based on experimental data from separate and integral test facilities. The TRACE is a member of the western PWR codes family which validation process is mainly based on the data from experimental facilities or real NPPs of western PWR type. There is a significant number of LWR NPPs which utilize VVER design of reactors operating all over the world and many other are under construction or planned as well. The VVER design of LWR is generally similar to western PWRs but there are several features that should be taken into consideration such as larger volumes of primary coolant, horizontal steam generators, different ECCS injection points and so on. While a lot of experimental data measured on PWRs or PWR-based test facilities are valuable also for research of VVER design, the specific validation of thermo-hydraulic codes for VVER design reactors is often required by national regulators. The purpose of performed analyses is to extend the validation of the TRACE code focused on VVER type of NPPs. The best estimate thermo-hydraulic computer code TRACE V5.0 patch03 and patch04 was assessed using Cold Leg 4.1%

break experiment at the large-scale test facility PSB-VVER.

The PSB-VVER facility is a 1:300 volume-power scaled model of VVER 1000 NPP located at Electrogorsk, Russia. In order to perform code validation an extensive TRACE input deck of PSB-VVER facility was developed. The TRACE model includes all important components of the PSB-VVER facility: reactor, four separated loops, pressurizer, break units, main circulation pumps, steam generators, and important parts of secondary circuit.

The main goals of the experiment 4.1 % cold leg break (CL-4.1-03) were as follows:

• to study VVER-1000 thermal hydraulics in an accident caused by a small break in the cold leg

• to investigate scale effect

• to investigate design effect of the horizontal SG tube arrangement

• to provide data for code assessment regarding the prediction of a VVER system thermal hydraulic response following the small break loss-of-coolant accident The test was also intended to be similar to the experiments performed in western integral test facility LSTF (test SB-CL-21), BETHSY (test 6.2 TC), LOBI (test BL-34) and SPES (test SP-SB-03). The time course of the transient that was performed in the LOBI facility was replicated in the test performed in the PSB-VVER facility and may be divide into four main stages:

• blowdown of subcooled coolant and the first heating of fuel assembly

• decreasing the primary coolant mass at high pressure and re-flooding after actuation of the ECCS ACCs

• decreasing the primary coolant mass at low pressure after stopping the operation of the ECCS ACCs and the final heating of the fuel assembly model

• actuation of the LP ECCS and the flooding of the primary circuit 1-1

2 FACILITY AND TEST DESCRIPTIONS Detail information about the PSB-VVER test facility systems and elements is given in Ref 11.

Only a brief description of the PSB-VVER facility is given here. The hardware configuration for CL-4.1-03 test is reported below.

PSB-VVER Facility PSB-VVER is a large-scale integral test facility which structurally corresponds to primary circuit of NPP with VVER-1000 (V-320 design). The volumetric and power scale is 1:300, and the main equipment elevations correspond to those of the prototype reactor.

The facility consists of four loops linked up to the reactor model. Each loop has a circulation pump, a steam generator, hot and cold legs. One of the loops (loop No.4, "broken") has special branch pipes for connection to primary leakage simulation system. The test facility also includes a pressurizer (PRZ) and ECCS, which has, as in actual VVER-1000, three subsystems: a passive system and two active ones.

The reactor model comprises four elements: an external downcomer, core model, core bypass and an upper plenum. The PSB-VVER core model consists of 168 full-height indirectly electrically heated fuel rod simulators with uniform power distribution. The rod simulator pitch (12.75 mm) and diameter (9.1 mm) are identical to those of the reference reactor. The fuel rod simulators are arranged on a triangular grid. The rod bundle cross section has the shape of regular hexagon with "wrench" size of 168 mm. The core model represents the central part of the reference fuel rod assembly. Along its height, the assembly has 15 spacer grids with natural geometry PSB-VVER pressurization system includes a pressurizer, surge lines, spray lines, and a relief valve. By means of surge and spray lines the pressurizer can be connected to the broken loop (loop #4) or to one of the intact loops (loop #2) of the facility. The PRZ vessel height, the bottom elevation and location of nominal level correspond to the reference ones. An electric heater with a power of up to 80 kW is built in the lower part of the pressurizer vessel.

PSB main circulation pumps are used to provide forced circulation in primary circuit. The circulation pumps are variable-speed ones of vertical centrifugal single-stage type and can operate under two-phase fluid conditions.

The passive ECCS system consists of four accumulators connected in pairs to an inlet and outlet chamber of the reactor pressure vessel. The active ECCS system consists of high pressure injection system (HPIS) and low pressure injection system (LPIS). Cooling water of active ECCS can be supplied to three loops, both to cold and hot legs as original facility design.

The PSB-VVER SG is a vertical vessel with two vertical headers inside. A bundle of horizontal spiral heat-exchanging tubes of full size is mounted between the two headers. The PSB-VVER SG is designed in such way that the reference tube bundle elevations and tube lengths to be 2-1

conserved, as well as the flow area. Heat transfer surface and secondary fluid volume to be matched the scale factor. On the secondary side, the feed water system and the main steam lines are simulated. The turbine and the condenser are not modeled.

Primary and secondary circuits of the PSB-VVER facility are operated at nominal pressure of a reactor prototype.

Figure 1 depicts an isometric projection of the test facility. Main operational characteristics of the test facility are given in Table 1.

Figure 1 General View of PSB-VVER Facility 2-2

Table 1 Main Operational Characteristics of PSB-VVER Comparing to VVER-1000 Parameter Units VVER-1000 PSB-VVER Coolant - water water Number of circulation loops - 4 4 Primary circuit Pressure MPa 15.7 15.7 Coolant temperature (hot/cold leg) deg 290/320 290/320 Coolant flowrate m3/h 82485 < 280 Core power MW 3000 15 Secondary circuit Steam generator pressure MPa 6.3 6.3 Feed water temperature deg 220 < 270 Thermal power of one SG MW 750 2.5 The PSB test facility is equipped with several special break systems to facilitate research of thermal hydraulics during break accidents. There is a special system to simulate:

• accumulator water supplying pipe rupture (11% upper plenum break experiments UP-11-07, UP-11-08)

• guillotine break of hot leg (2x25% hot leg large break experiment HL-2x25-02)

• guillotine break of cold leg (100 % double ended cold break experiment CL-2x100-01)

• small cold leg break (4.1 % cold leg break test CL-4.1-03)

• primary to secondary leak (PRISE test Psh-1.4-04) 2-3

Experiment CL-4.1-03 The test CL-4.1-03 4.1 % Cold Leg Break test was performed in the PSB-VVER test facility at Electrogorsk Research and Engineering Center (EREC) in Russia. The thermal-hydraulic processes related to cold leg break 4.1% were investigated.

Facility Configuration The information on the test facility hardware and configuration of the system specific for CL-4.1-03 test is given in the Table 2.

Table 2 Test Facility Configuration in CL-4.1-03 Test Equipment Status Pressurizer Connected to the loop #4 Core by-pass 2 diaphragms with 2 orifices with diameter of 7 mm are installed at inlet and outlet of core by-pass HPIS Not used LPIS ECCS water is injected in cold legs of loops # 1, 3 and 4 ACCs Two accumulators are connected to the inlet chamber SGs Under steady state all SGs are connected to each other by steam header. The pressure is adjusted by one steam dumping valve RA06S01 ADS ADS (atmospheric steam dump system) is connected to each SG. In each ADS line, throttle channel (L/d=10) 12.1 mm in diameter is installed Feed water heater Under steady-state conditions the SG levels are maintained by means of impulse supply of the feed water with temperature ~

170 C°.

Large break unit Upward vertically oriented break line is connected to loop #4 cold leg between MCP-4 and pressure vessel inlet. The leak is limited by a smoothly edged (R=6 mm) break nozzle 10 mm in diameter and 100 mm length.

UP warming-up line Under steady state the line is open. The warming-up of the UP top part is stopped about 2 min before opening break line.

Warming-up line of Under steady state the line is open. The warming-up of the break line break line is stopped about 1 min before opening break line.

Initial Conditions The main initial conditions of CL-4.1-03 test are given in the Table 3. The CL-4.1-03 test has been performed under reduced initial core power corresponding to approximately 15% of nominal power.

Table 3 Measured Initial Condition for CL-4.1-03 Test Parameter Units Value Primary circuit Pressure in upper plenum (gauge YC01P16) MPa 15.512 Coolant temperature deg C 282.4 / 309.2 2-4

Table 3 Measured Initial Condition for CL-4.1-03 Test (Cont'd)

(DC inlet/UP outlet - gauges C01T02/YC01T04)

Primary loops flow rates (gauges YA01÷04F01) kg/s 1.964 / 1.989 / 1.979 / 1.937 Core power (gauge YC01N01) kW 1128.6 Core by-pass power (gauge YC01N02) kW 14.9 Coolant level in PRZ (gauge YP01L02) m 3.073 Secondary circuit Pressure in SGs (gauges RA01÷04P01) MPa 6.878 / 6.913 / 6.936 / 6.886 Level in SGs (gauges YB0104L01) m 1.904 / 1.913 / 1.935 / 1.901 ECCS Pressure in ACCs (gauges TH02,04P01) MPa 4.079 / 4.072 Level in ACCs (gauges TH02,04L01) m 4.577 / 4.570 Boundary Conditions (Test Scenario)

Detail information about the CL-4.1-03 test boundary conditions is given in Ref 14. The main events of CL-4.1-03 test are described in the Table 4.

Table 4 Main Events During CL-4.1-03 Test Event Time [s]

Cutting off electric load at PRZ -0.6 IE - Break Opening 0 Pressure in UP (YC01P16) < 13.73 MPa conditions for SCRAM 4.1 Stopping of MCPs 1 ÷ 4 () 7.8 Pressurizer emptying(YP01L02) 10 Feed water valve closure (RL01÷ 04S06) 14 Main steam valve closure (RA06S01) 18 Core and core by-pass power reduction onset (YC01N01, 02) 57.6 Accumulators injection onset : TH02B01 / TH04B01 414 / 406 Accumulators injection termination (L < 0.559 m): TH02B01 /

1452 / 1365 TH04B01 LPIS injection onset (TJ01F02 / TJ03F02 / TJ03F02) 2432 / 2432 / 2434 End of test (FRS power switched off) 2593.4 The experiment is started with opening the break valve XL01S01, after that the PSB-VVER PCS gives signal to disconnect the PRZ heaters.

The actions on reaching the primary pressure of value 13 MPa are as follows:

• closure of secondary pressure control valve RA06S01, while the steam generators remain connected with each other by the steam line;

• closure of valves RL01-04S06 and cutoff feed water into steam generators;

• trip of all MCPs, the pump rotor speed drops from its initial value to zero in 4 s

• core power follows the power/time relation given in following table:

2-5

Table 5 Time Dependence of Dimensionless Core Power Time (s) Power/Initial power (-) Time (s) Power/Initial power (-)

0 1 400 0.224 53 1 600 0.196 60 0.941 800 0.186 80 0.783 1000 0.178 100 0.662 1500 0.163 150 0.464 2000 0.148 200 0.359 2375 0.144 When primary pressure drops below 4 MPa, the ECCS accumulators start to inject water in the pressure vessel inlet chamber. Accumulators remain connected to the inlet chamber until the water levels decrease below 1.31 m (0.559 as readings TH0204L01). Then the accumulators are disconnected by valves TH02S05 and TH04S05 in order to avoid air penetration into the primary system.

The low pressure ECCS is activated at the FRS temperature 500 °C, provided that the accumulator water injection is already terminated and the tank are disconnected from the primary system. The water is supplied in cold legs of loops #1, 3 and 4 with the mass flow rate 0.248 kg/s in each line, the flow rate is regulated by the valves TJ01S08, TJ03S08 and TJ04S08.

The SG ADS valves open at secondary pressure 7.4 MPa and close at 7.2 MPa. The open/close set points of the ADS valves are derived from the plots demonstrating the secondary pressure behavior in LOBI facility. Throughout the transient, the steam generators remain connected to a common steam header; therefore the steam generators secondary pressures are expected to be equal. But an insignificant uncertainty in measuring the pressures can lead to opening ADS valve in some steam generator while the others remain closed due to decrease in the secondary pressure. Therefore, all the ADS valves are assumed to be dependent on SG-1 pressure (YB01P01).

The experiment had to be ended on reaching the FRS temperatures to 700 °C or on reaching the steady core cooling conditions after the final core rewet.

2-6

3 THE TRACE V5.0 CODE The TRACE (TRAC/RELAP Advanced Computational Engine) is the latest in a series of advanced, best-estimate reactor systems codes developed by the U.S. Nuclear Regulatory Commission for analyzing transient and steady-state neutronic-thermal-hydraulic behavior in light water reactor.

The TRACE code has been widely used by U.S. Nuclear Regulatory Commission (NRC) and other organizations for rulemaking, licensing audit calculations, evaluation of operator guidelines, and as a basis for a nuclear plant analyzer. Specific applications of their capability have included simulations of transients in LWR systems, such as loss of coolant, anticipated transients without scram (ATWS), and operational transients such as loss of feedwater, loss of offsite power, station blackout, and turbine trip. The TRACE is a highly generic code that, in addition to calculating the behavior of a reactor coolant system during a transient, can be used for simulation of a wide variety of hydraulic and thermal transients in both nuclear and nonnuclear systems involving mixtures of steam, water, noncondensable gases, and solute.

The TRACE code is a component-oriented reactor systems analysis code designed to analyze light water reactor transients up to the point of significant fuel damage. The TRACE code solves a finite-volume two-phase multidimensional compressible flow with one, two and three dimensional flow geometry. The TRACE code can model heat structures and control systems that interact with component models and the fluid solution. The TRACE code has capability to use build-in point reactor kinetics or 3D reactor kinetics through coupling with Purdue Advanced Reactor Core Simulator (PARCS). In addition the TRACE code can be coupled with another TRACE jobs or other codes (CFD, CONTAIN ...) through its exterior communications interface (ECI). TRACE uses what is commonly known as a 6-equation model for two-phase flow (mass equation, equation of motion and energy equation for each phase). Additional equations can be solved for noncondensable gas, dissolved boron, control systems and reactor power. There are five additional closure relationships for field equations: equations of state, wall drag, interfacial drag, wall heat transfer and interfacial heat transfer. These constitutive models are semi empirical equations. There are two numerical methods available in TRACE: semi-implicit method and the stability enhancing two-step (SETS) method.

3-1

The TRACE V5.0 Code Assessment Confidence in the computational tools (codes) and establishment of their validity for a given application depends on proper assessment. The TRACE code, like other two-fluid codes, is composed of numerous models and correlations. When applied to full scale nuclear power plant conditions, many of these models and correlations can be applied outside of their original scope. By assessing the code against thermal-hydraulic tests, it is possible to show that the code and its constituent model packages can be extended to conditions beyond those for which many of the individual correlations were originally intended (Ref 4). The assessment process however, can also indicate potential deficiencies in the code. There are following four sources of data for code assessment (Ref 21):

• Fundamental experiments

• Separate effect test facilities (SETF)

• Integral test facilities (ITF)

• Real plant data VVER Typical Features Related to TRACE V5.0 Code Assessment The TRACE code validation process is mainly based on the data from experimental facilities or real NPPs of Western PWR type. VVER reactors are in many aspects similar to Western PWRs.

Therefore a lot of experimental data measured on PWRs or PWR test facilities is valuable also for VVER research. On the other hand, the VVER design has several specific features. From the hardware point of view the main differences between VVER-1000 and PWR are the following (Ref 9):

• Horizontal steam generators with 2 headers

• Lower plenum internal structures

• Fuel assemblies with hexagonal fuel rod arrangements

• ECCS injection points

• Secondary side water volume of the steam generators is larger

• Operational conditions and set points of actuation of ECCS

• Working conditions of secondary side of steam generators and set points for the operation of feedwater and steam lines There are approximately 60 operating units of VVER type (Ref 20). It is a meaningful number in comparison to approximately 216 operating units of PWR reactors (Ref 20). Therefore corresponding attention should be given to code validation for VVER type of reactors.

3-2

4 INPUT DECK DESCRIPTION An extensive TRACE input deck of PSB-VVER facility was developed including all important components of the PSB-VVER facility: reactor, 4 separated loops, pressurizer, break units, main circulation pumps, steam generators, break sections and important parts of secondary circuit. The input decks were designed on the basis of PSB-VVER facility documentation (Ref 11, 12).

The TRACE Input Deck Nodalization diagrams of the TRACE Input Deck are presented in Figure A-1 (reactor + primary circuit) and Figure A-2 (secondary circuit) in the Appendix A. The TRACE model of the reactor pressure vessel (RPV) with internal structures is divided into 3 parts. The first part represents Downcomer (DC) + Lower Plenum (LP), the second part represent Fuel Rod Simulator (FRS) +

Upper Plenum (UP) and the third part represents core by-pass from DC to UP. The RPV model employs VESSEL component for DC + LP (includes 26 axial layers, 1 azimuthal theta sector and 2 radial rings), 3 PIPE components for core by-pass and the next VESSEL component for FRS + UP (includes optionally 25 or 26 axial layers, 4 azimuthal theta sector and 3 radial rings).

The model also includes the by-pass piping between DC to UP and the UP heating pipe between cold leg of loop #1.

Each of the four coolant loops comprises of: a hot leg, steam generator, pump suction loop seal piping, main coolant pump, and a cold leg including control valve between MCP discharge and DC.

The pump performance is based on single-phase head and torque characteristic of the pump TsNIS 1620 from Ref. 12. No two-phase degradation was modeled because no appropriate data are available.

The pressurizer is modeled using component PIPE equipped with heaters and with surge lines connected to the hot legs of loop #2 and loop #4. Pressurizer can be optionally connected to loop #2 or loop #4 (as original facility design).

Active ECCS are modeled using simple boundary condition - FILL component, accumulators are modeled using PIPE and VALVE components. High pressure injection system (HPIS) optionally provides flow to the hot leg of loop #1. Low pressure injection system (LPIS) optionally provides flow to the hot leg of loop #1 and #3 and to the cold leg of loop#1, loop#3 and loop#4. Four accumulators provide flow to the downcomer and upper plenum (two to each location), and any of them can be switch off. Cooling water delivery from ECCS depends on hardware configuration of a particular test.

Steam generator is modeled using multi-tube approach. The primary side of SG Input Deck consists of 5 axial layers of heat exchanging tubes and two headers (the original facility SGs consists of 34 tubes). Each axial layer is divided into 15 segments. The SG secondary part is modeled as an original three-channel complex with 10 axial layers (5 of them in the area of 4-1

exchanging tubes). The feedwater system, the steam lines connected to all SGs and the common steam header are also modeled. The BREAK component simulates the release of secondary steam from steam header. The secondary side of all SGs is equipped with the ADS (atmospheric dump system) using VALVE and BREAK component.

The heat loss from the primary and secondary is represented in the TRACE model by entering the thickness of the insulation on the outer surface of all the pipes and other system components. Appropriate material properties are input for the insulation. A constant boundary temperature and heat transfer coefficient of outer air is applied.

The model contains 1239 volumes, 2176 junctions, and 1114 heat structures with 4174 mesh points. Standard modeling guidelines were followed in developing the nodalization of the system.

Components Statistic for TRACE model - see the next Table 6.

Table 6 TRACE Components Statistic TRACE Component Notes VESSEL 2 DC+LP; FRS+UP PIPE 67+66*1 -

HSTR 139+1*1 -

POWER 1+1*1 FRS simulator + By-pass heating VALVE 30 -

PUMP 4 MCPs Upper plenum, Large and Cold Leg Break BREAK 8 units, release of secondary steam, SG ADS FILL 10 HPIS, LPIS, Feedwater, Make-up system Total No of Components 330 -

• 1 the second number is the number of spawned component 4-2

Table 7 List of the Main Systems and Components of PSB-VVER TRACE Input Deck PSB system Input deck Used TRACE components Rector (YC)

Downcomer + VESSEL + HTSTR Lower plenum + VESSEL + PIPE + HTSTR Core + VESSEL + HTSTR + POWER + CONTROL BLOCK Upper plenum + VESSEL + HTSTR Core bypass + PIPE + HTSTR + POWER + CONTROL BLOCK DC to UP bypass + VALVE + HTSTR UP heating + VALVE + HTSTR LOOP (YA)

Hot leg + PIPE + HTSTRT Loop seal + PIPE + HTSTRT Cold leg + VALVE + HTSTRT Main cooling pump (YD) PUMP + HTSTR + CONTROL BLOCK Pressurizer (YP)

Vessel + PIPE + HTSTR heaters + HTSTR + CONTROL BLOCK Surge line + VALVE + HTSTR Relief valve -

Make-up system (+) FIL + CONTROL BLOCK ECCS (TJ, TH)

HPIS (+) FILL + CONTROL BLOCK LPIS (+) FILL + CONTROL BLOCK Accumulators + PIPE + VALVES + HTSTRT Steam generators (YB)

Vessel + PIPE + HTSTR Heat exchange tubes + PIPE + HTSTR Primary headers + PIPE + HTSTR Steam lines + VALVE + HTSTR Feedwater (+) FILL + CONTROL BLOCK Relief valve -

ADS (+) VALVE + BREAK + CONTROL BLOCK Steam headers (RA) + VALVE + HTSTR + BREAK Key: + a fine model

(+) a simplified model

- not modeled 4-3

5 RESULTS Steady-State Calculation In order to achieve stable initial conditions of the CL-4.1-03 test, the steady state was calculated for 1200 s. The following controllers were used for the first 1200 s:

• Pressurizer pressure controller

• Pressurizer level controller

• Steam generators level controllers The other controlled parameters (fuel rod simulator power, core bypass power, feedwater temperature, main steam header pressure) were entered as boundary conditions. The steam generator pressures are lower than the measured values, because the steam header pressure was adjusted to get the desired reactor vessel inlet temperature (average cold legs temperature) in steady state calculation. A real PSB controller that kept the liquid level in all SGs within a desired band was replaced by a PI-controller for 1000 s of steady-state calculation, for the rest of steady state calculation and for transient calculation was feed water flow to SGs entered as a boundary condition. Reaching steady state conditions took approximately 41 min.

Main calculated and measured parameters are compared in the following Table 8.

Table 8 Initial Conditions (TRACE Calculation vs. Experiment Comparison)

Parameters TAG Units accuracy CL-4.1-03*2 TRACE Primary Circuit Upper plenum pressure YC01P16 MPa +/- 0.05 15.512 +/- 0.004 15.513 YC01DP07-Pressure drop at FRS kPa +/- 2.40*1 -28.6 +/- 0.021 -29.4 DP10 Upper Plenum Coolant YC01T04b °C +/- 4.2 311.5 +/- 0.1 308.9 temperature Core Inlet Coolant Temperature YC01T259 °C +/- 3.0 282.5 +/- 0.1 282.7 Hot Leg outlet Coolant Temp. TA01T03 °C +/- 3.0 310.4 +/- 0.1 308.3 Hot Leg outlet Coolant Temp. TA02T03 °C +/- 3.0 308.1 +/- 0.4 309.4 Hot Leg outlet Coolant Temp. TA03T03 °C +/- 3.0 310.5 +/- 0.1 309.4 Hot Leg outlet Coolant Temp. TA04T03 °C +/- 3.0 308.5 +/- 0.4 309.6 Cold Leg outlet Coolant Temp. TA0T02 °C +/- 3.0 283 +/- 0.1 282.7 Cold Leg outlet Coolant Temp. TA02T02 °C +/- 3.0 283.5 +/- 0.1 282.7 Cold Leg outlet Coolant Temp. TA03T02 °C +/- 3.0 282.4 +/- 0.1 282.7 Cold Leg outlet Coolant Temp. TA04T02 °C +/- 3. 281.8 +/- 0.1 282.7 Loop-1 flow rate YA01F01 kg/s +/- 0.1 1.964 +/- 0.012 1.929 Loop-2 flow rate YA02F01 kg/s +/- 0.1 1.989 +/- 0.007 1.948 Loop-3 flow rate YA03F01 kg/s +/- 0.1 1.979 +/- 0.007 1.945 Loop-4 flow rate YA04F01 kg/s +/- 0.1 1.937 +/- 0.013 1.913 FRS power YC01N01 kW +/- 15 1128.6 +/- 8.3 1128.6 Core by-pass power YC01N02 kW +/- 0,4 14.87 +/- 0.02 14.9 Collapsed level in PRZ YP01L02 m +/- 0.3 3.073 +/- 0.017 3.073 5-1

Table 8 Initial Conditions (TRACE Calculation vs. Experiment Comparison) (Cont'd)

Parameters TAG Units accuracy CL-4.1-03*2 TRACE Secondary Circuit Pressure in SG-1 YB01P01 MPa +/- 0.05 6.878 +/- 0.002 6.784 Pressure in SG-2 YB02P01 MPa +/- 0.05 6.913 +/- 0.002 6.783 Pressure in SG-3 YB03P01 MPa +/- 0.05 6.936 +/- 0.002 6.776 Pressure in SG-4 YB04P01 MPa +/- 0.05 6.886 +/- 0.002 6.785 Collapsed level in SG-1 YB01L01 m +/- 0.08 1.904 +/- 0.002 1.904 Collapsed level in SG-2 YB02L01 m +/- 0.08 1.913 +/- 0.002 1.913 Collapsed level in SG-3 YB03L01 m +/- 0.08 1.935 +/- 0.002 1.935 Collapsed level in SG-4 YB04L01 m +/- 0.08 1.901 +/- 0.003 1.901 Accumulators ACCU-2 pressure TH02P01 MPa +/- 0.03 4.577 +/- 0.001 4.577 ACCU-4 pressure TH04P01 MPa +/- 0.03 4.570 +/- 0.001 4.570 ACCU-2 collapsed level TH02L02 m +/- 0.07 4.079 +/- 0.000 4.080 ACCU-4 collapsed level TH04L02 m +/- 0.07 4.072 +/- 0.001 4.074

• 1 sum of accuracy of pressure drop YC01DP07-DP10 (accuracy of YC01DP07,08,10 =

+/- 1.2 kPa, YC01DP09 = +/- 1.0 kPa)

• 2 Average value +/-standard deviation of measured parameters at initial steady state condition of the test facility 5-2

Transient Calculation The post-test calculation of the CL-4.1-03 experimental test at PSB-VVER test facility started at the time 0 s simultaneously with an initiating event - Cold Leg break 4.1 % and disconnecting of the PRZ electric heater. A comparison of calculated and experimental times of the occurrence of main events is presented in the Table 9. Time courses of all important parameters and their comparison with experimental data are presented in Appendix C.

A brief overview of the behavior observed in the experiment is provided here, then the comparisons between the measured data and calculations will be presented and discussed.

Time Course of the Transient The time course of the transient can be divided into four main stages:

• blowdown of subcooled coolant and the first heating of fuel assembly

• decreasing the primary coolant mass at high pressure and re-flooding after actuation of the ECCS ACCs

• decreasing the primary coolant mass at low pressure after stopping the operation of the ECCS ACCS and the final heating of the fuel assembly model

• actuation of the LP ECCS and the flooding of the primary circuit Immediately after the leak flow onset, the primary pressure rapidly decreased (Figure C-1). At the time 4.1 s the primary pressure decreased down to 13.0 MPa followed by closing RA06S01 valve in steam dumping line and RL01-04S06 valves in feed water lines. At the same time all MCPs were also turned off and they were fully stopped by 8 s. Power loss at the assembly model and by-pass began at 57 s. By 1620 s the power at the fuel rod simulator (FRS) was decreased down to 180 kW and later on it has constant value (Figure C-7).

As a result of the primary pressure decrease, the PRZ coolant boiled up and was completely displaced into hot leg of loop 4. By the 10 s, the PRZ fluid level reached the lower pressure tap elevation mark (Figure C-3, gauge YP01L02).

Mass-flow rates through the circulation loops during first 160 s are presented in Figure C-27.

The pressure drops gauge YA01-04DP04 was utilized to show the collapsed level in cold legs above flow meter locations (Figure C-15, C-17, C-19 and C-21). The absolute value of pressure drops close to zero value indicates that after 120 s the collapsed level in cold legs is below the flow meter locations. After that time the YA01-04F01 measurements are not valid.

After the steam dump line valve was closed at the 18 s, there took place rapid growth of pressure (Figure C-6) caused by heat transfer from primary circuit into secondary circuit. The growth of secondary pressure continued until it reached the setpoint and actuated atmospheric steam dump system. The signal to open the valves simulating the ADS operation is conditioned by the SG1 pressure transducer (YB01P01). The valves RA11-14S01 simulating the ADS operation are opened three times in total. The ADS system stopped acting after 150 s, when the primary pressure fell down below the secondary one and reverse heat exchange through SGs took place (SG did not remove heat from primary circuit any more).

On reaching the coolant saturation point and onset of its sub-boiling, phase separation is begun in the primary circuit. Boling of primary coolant was followed by formation of a steam 5-3

bubble in the upper part of UP at 16 s, in the SG hot headers at 20 s and in the SG cold headers at 36 s. Hotter water discharged from PRZ to the hot leg of loop#4 reached saturation point 7 s earlier than in other steam generators.

Simultaneously with drying of the upper plenum, the levels in the SG headers in the cold and hot legs decreased as well (Figure C-15, C-17, C-19 and C-21). The drying of the SG headers resulted in the primary coolant circulation ending in loop 4 at the 55 s and in other loops at the 80 s (Figure C-27).

As the result of circulation ending in primary loops, there begun intensive flashing of the core model (Figure C-22). By the 100 s the core coolant mass was decreased down to minimal value. The first heating of the fuel rod simulator cladding surface in the top part of the assembly occurred at the 97 s, it was an consequence of that core coolant mass decrease (Figure C-2).

The heating of the assembly was stopped at the 105 s as the result of loops 1 and 4 seals water clearance (Figure C-16 and C gauge YA01,04DP05,06) into downcomer and then into the core model. Clearance of loops 1 and 4 seals were realized completely by the 300 s. Since the 143 s, there was observed presence of steam in the break. This is evident from specific increasing rate of primary pressure reduction (Figure C-2). Since that moment, the primary circuit did not need in steam generators to remove heat and as result of that, there took place reverse of heat exchange between the primary and secondary circuits.

At the 113 s, the coolant in downcomer reached state of saturation as the result of loop seal clearance. By the 124 s the level in downcomer was lowered down to cold leg elevation marks and slowly continued decreasing till the 227 s (Figure C-23). After the 113s, there occurred discharge of the saturated coolant through the break.

The upper plenum was completely dried by the 380 s. After that the core region drainage started (Figure C-22). The second heating of the fuel rod simulator cladding surface begun at the 405 s as the result of their top part dry-out (Figure C-2). The heating was stopped after the primary pressure decreased below pressure in accumulators and begun their discharge into the downcomer.

Actuation of the accumulators TH04B01 and TH02B01 took place at the 406 and 414s respectively (Figure C-5 and C-6). The water from the accumulators TH02B01 and TH04B01 is entering downcomer inlet of the reactor model. Emptying the accumulators and disconnection them from the primary circuit took place at the 1365 s (TH04B01) and at the 1452 s (TH02B01).

As both accumulators were supplying water into the same place of the primary circuit, difference in their disconnection time may be explained due to different hydraulic resistances of their connections.

As the result of accumulators operation there took place raising of levels in the reactor model within period of the time 420-800 s (Figures C-23 to C-25). During accumulators discharge the level in the reactor model was kept constant. After termination of operation of accumulators there begun drainage of the upper plenum and downcomer again. At the 1975 s (when the upper plenum was dried almost completely) the level in the fuel core simulator began to decrease. The third heating-up of the fuel rod simulator cladding surface begun at the 2057 s (Figure C-2). It was continued up to the LP ECCS actuation (at the 2432 s) by the signal "overshoot of the FRS cladding surface temperature" (up to 500 °C) 5-4

After LP ECCS water supply started with flow rate 0.248 kg/s into cold legs of loops #1, 3 and 4, rapid flooding of the primary circuit begun. By the 2570 s, temperature of simulated fuel element surfaces decreased down to saturation point, and at the 2593 s the experiment was stopped.

Comparison of TRACE Prediction to Reference Data Comparison of the calculated and measured results showed that calculated values of primary pressure were in a very good agreement with experimental data (Figure C-1). But initial calculated rate of the primary pressure decrease was considerably higher so the set-point 13.0 MPa was reached earlier at 2.2 s unlike in the experiment at 4.1 s. During the time period from the reversal of heat exchange from primary and secondary circuits to the start of the accumulators discharge the calculated primary pressure was decreasing more rapidly than the pressure in the experiment. As the result, the set point of the accumulators is reached sooner (Figure C-6).

To properly predict fuel cladding temperature peaks the option Level tracking was activated in the core and the UP region. As a result of that the accumulators discharge turned to be stepwise. The stepwise discharge was caused by temporally increasing pressure in the core after water from accumulators had been delivered to the core evaporated.

The TRACE code predicted a little bit slower decreasing liquid water in the upper part of the core during the first stage of the experiment so the first temperature peak between the 97 s and 105 s was not predicted. The absolute temperature and time occurrence of the second and the third temperature peak were predicted very well (Figure C-2).

Water distribution (pressure differences) over the reactor vessel and circulation loops (Figure C-15 to C-26) was in a good compliance with experiment data. The decreasing of the absolute value of the measured pressure difference means drainage of a particular region. There are only two regions where the TRACE code predicted different liquid content: the middle part of the core and loop seals (a part including the MCP and the filter). The calculated liquid content of the middle part of the core was somewhat lower than measured one (Figure C-24 and C-25 gauge YC01DP07, 08, 09). The calculated rate of clearance of the loop seals of loops with no break was slower in case the loop#1(Figure C-16) and faster in case the loops #2 and #3 (Figure C-18 and C-20). But these discrepancies did not significantly affected predicted heat-up of the upper part of the core, where it was expected to happen first (Figure C-2). The third temperature peak overcame the FRS cladding temperature set-point 500 °C nearly at the same time in the experiment and calculation followed by acting of LPIS ECCS.

The time course of the secondary pressure including pressure peaks before ADS was activated is in very good agreement with experimental data.

Table 9 Chronology of Main Events (TRACE Calculation vs Experiment Comparison)

Time [s]

Event CL-4.1-03 TRACE Cutting off electric load to PRZ heaters -0.6 0 Initiating Event - break opens 0 0 Pressure in UP (YC01P16) < 13.70 MPa - SCRAM signal 4.1 2.2 5-5

Table 9 Chronology of Main Events (TRACE Calculation vs Experiment Comparison) (Cont'd)

Stopping of all MCPs 7.8 6 Pressurizer emptying (YP01L02) 10.0 9.0 Feedwater valves closure 14 14 Saturation condition at core outlet 16 20 Main steam valve closure 18 3.

Start of core and core by-pass power reduction 57.6 55.2 First core dry-out (peak cladding temperature) 97 -

Two-phase flow break 113 135 Primary to secondary pressure reversal 150 159 Second core dry-out (peak cladding temperature) 405 358 Start of ACCU-2 injection 414 342 Start of ACCU-4 injection 406 343 End of ACCU-2 injection 1452 1848 End of ACCU-4 injection 1365 1600 Final core dry-out (peak cladding temperature) 2057 2218 Start of LPIS injection 2432 2408 End of test 2593 2600 Sensitivity Studies Performed Before final calculation a complex sensitivity study of the impact of various parameters and code options on the prediction of the rod heaters cladding temperature was performed. The parameters with no impact on prediction of the rod heaters cladding temperature are as follows with short explanation:

• CCFL at core outlet plate: The accumulators were connected to the DC and ECCS LPIS were connected to the cold legs so core was flooded from the bottom to the top and such a phenomena as CCFL did not occur.

• Offtake model at the break (the top and bottom orientation of the break): No horizontal stratified flow occurs in the pipe with the break.

• Discharge coefficients at the break: No choking occurred during time period when the TRACE over predicted the break flow (higher the primary pressure rate before accumulators operation).

The parameters affected prediction of the rod heaters cladding temperature are as follows with short explanation:

• The Level Tracking model in the core and UP region.

• The finer discretization of the upper part of the core

• To utilize the latest version TRACE V5.0 patch 04 (at the time when calculations were performed)

The activation of the Level tracking model introduced the presence of the second temperature peak. All other parameters improved time occurrence and absolute value of the second and the third temperature peak.

For the final calculations the finer discretization of the upper part of the core and Level tracking model in the core and UP were applied. The calculation is optionally performed using the TRACE code V5.0 patch03 and patch 04 to access changes between versions of the TRACE ode. Comparison of the test data and calculated results is presented in the appendix C.

5-6

Measured data has a circle symbol and calculated results are labeled p03 for TRACE V5.0 patch03 results and p04 for TRACE V5.0 patch04.

Quantitative Assessment of the Calculations To quantify agreement of presented TRACE calculations the figure of merit (FOM) was evaluated using software ACAP (Automated Code Assessment Program), which is a part of the software package SNAP. Settings of ACAP was based on Ref 22 including choice of particular metrics and their weighting factors - and see the Table 10.

Table 10 ACAP Metrics Settings Metric name Abbreviation Weighting factor D'Auria Fast Fourier Transformation FFT 0.35 Mean Error Magnitude MEM 0.35 Size-Independent (Pred - Perf) Norm SI-PMPN 0.15 Degree of Randomness DOR 0.15 To assess the value of FOM, acceptability criterions were established on the basis of Ref 23, where the FFTB method (Fast Fourier Transform Based Method) is described. FOM acceptability criterions were based on AAtot (total average amplitude). Value of AAtot is transformed to FOM using the equation of D'Auria FFT metric:

1

= 1 2 2

([2 + ( ) ] + 1)

Where k is weighted frequency importance factor and value k = 0 was applied, which means that pure magnitude error is evaluated using D'Auria FFT metric. The next Table 11 contains values of acceptability criterions range and their meaning.

Table 11 Acceptability Criterions AAtot range FOM range Abbreviation Color indication AAtot 0.30 FOM 0.77 Very good code predictions green 0.30 < AAtot 0.50 0.67 FOM < 0.77 Good code predictions blue 0.50 < AAtot 0.70 0.59 FOM < 0.67 Poor code predictions orange AAtot > 0.70 FOM < 0.59 Very poor code predictions red To assess TRACE and RELAP5 calculation, the representative set of 71 parameters were chosen including:

• Primary pressure

• Fuel cladding temperature

• Pressurizer water level

• RPV Pressure drops 5-7

• LOOPs pressure drops and mass flow rates

• Accumulator water levels and pressures

• Steam generators level and pressure

• LPIS mass flow rates To evaluate overall FOM uniform weighting factors were used for each of parameters.

The CL-4.1-03 experiment was a long lasting transient where many different TH phenomena occurred. In order to carefully assess both TRACE calculations the whole time course was divided into three time windows of interest as follows:

• W1: 0 ÷ 400 s - an early stage of the test when no ECCS flow was provided

• W2: 400 ÷ 1470 s - a middle stage of the test when accumulators flow was provided

• W3: 1470 ÷ 2593 s - the final stage of the test when ECCS LPIS flow was provided To assess the whole time course of the test FOM calculations with no time segmentation was performed as well. The following table contains all evaluated FOMs for all time windows of interest. To make results more readable color indication mentioned in the Table 11 was applied.

FOM = 1 means the best agreement and FOM = 0 the worst agreement. Location of PSB-VVER measurements is depicted in Appendix B for evaluated parameters.

5-8

Table 12 Evaluation of FOM for CL-4.1-03 Test TRACE V5 patch3 TRACE V5 patch4 all W1 W2 W3 all W1 W2 W3 Parameter TAG 0 - 2593 0 - 400 400 - 1470 1470 - 2593 0 - 2593 0 - 400 400 - 1470 1470 - 2593 Water level in accumulator of ECCS TH02B01 TH02L01 0.930 0.650 0.886 0.369 0.926 0.686 0.892 0.343 Pressure in accumulator TH02B01 TH02P01 0.893 0.467 0.851 0.422 0.901 0.488 0.867 0.431 Water level in accumulator TH04B01 TH04L01 0.917 0.742 0.865 0.370 0.923 0.769 0.874 0.474 Pressure in accumulator TH04B01 TH04P01 0.885 0.667 0.848 0.403 0.896 0.700 0.863 0.421 Flow rate in active ECCS system TJ01F02 0.737 0.752 0.706 0.710 0.794 0.752 0.706 0.789 Flow rate in active ECCS system TJ03F02 0.734 0.756 0.756 0.704 0.799 0.756 0.756 0.790 Flow rate in active ECCS system TJ04F02 0.737 0.738 0.743 0.710 0.790 0.738 0.743 0.782 Pressure drop in HL of loop 1 (elevation part) YA01DP02 0.771 0.791 0.708 0.726 0.771 0.794 0.709 0.727 Pressure drop in CL of loop 1 YA01DP04 0.857 0.840 0.616 0.567 0.878 0.862 0.617 0.569 (from SG outlet to the flow meter)

Pressure drop in CL of loop 1 YA01DP05 0.854 0.804 0.619 0.674 0.858 0.810 0.617 0.679 (from flow meter to mechanical filter)

Pressure drop in CL of loop 1 YA01DP06 0.706 0.662 0.313 0.583 0.699 0.664 0.304 0.556 (section with a mechanical filter, pump and regulating valve)

Pressure drop on the filter in CL of loop 1 YA01DP08 0.602 0.659 0.303 0.494 0.588 0.663 0.299 0.466 Pressure drop on MCP of loop 1 YA01DP09 0.715 0.657 0.277 0.482 0.705 0.658 0.267 0.432 Pressure drop in cold header of SG 1 YB01W01 YA01DP13 0.875 0.849 0.646 0.720 0.872 0.842 0.632 0.743 5-9 Pressure drop in hot header of SG 1 YB01W01 YA01DP14 0.889 0.895 0.665 0.644 0.890 0.896 0.665 0.648 Flow rate in the pump circulation loop YD01D01 (I 5) YA01F01 0.607 0.593 0.514 0.652 0.608 0.593 0.514 0.654 Pressure drop in HL of loop 2 (elevation part) YA02DP02 0.835 0.829 0.671 0.681 0.836 0.829 0.673 0.680 Pressure drop in CL of loop 2 YA02DP04 0.864 0.849 0.679 0.711 0.869 0.854 0.679 0.713 (from SG outlet to the flow meter)

Pressure drop in CL of loop 2 YA02DP05 0.667 0.676 0.588 0.513 0.667 0.677 0.588 0.517 (from the flow meter to mechanical filter)

Pressure drop in CL of loop 2 YA02DP06 0.703 0.818 0.661 0.668 0.714 0.820 0.671 0.679 (section with mechanical filter, pump and regulating valve)

Pressure drop on filter in CL of loop 2 YA02DP08 0.698 0.775 0.702 0.739 0.702 0.779 0.713 0.737 Pressure drop on MCP of loop 2 YA02DP09 0.689 0.828 0.658 0.523 0.701 0.827 0.664 0.537 Pressure drop in cold header of SG 2 YB02W01 YA02DP13 0.863 0.835 0.644 0.635 0.852 0.820 0.638 0.634 Pressure drop in hot header of SG 2 YB02W01 YA02DP14 0.917 0.903 0.725 0.729 0.922 0.908 0.734 0.728 Flow rate in the pump circulation loop YD02D01 (I 6) YA02F01 0.649 0.739 0.430 0.341 0.652 0.744 0.430 0.340 Pressure drop in HL of loop 3 (elevation part) YA03DP02 0.816 0.817 0.665 0.718 0.816 0.818 0.665 0.718

Table 12 Evaluation of FOM for CL-4.1-03 Test (Cont'd)

TRACE V5 patch3 TRACE V5 patch4 all W1 W2 W3 all W1 W2 W3 Parameter TAG 0 - 2593 0 - 400 400 - 1470 1470 - 2593 0 - 2593 0 - 400 400 - 1470 1470 - 2593 Pressure drop in CL of loop 3 YA03DP04 0.869 0.861 0.726 0.756 0.863 0.853 0.727 0.755 (from SG outlet to the flow meter)

Pressure drop in CL of loop 3 YA03DP05 0.658 0.671 0.603 0.635 0.656 0.673 0.604 0.629 (from the flow meter to mechanical filter)

Pressure drop in CL of loop 3 YA03DP06 0.695 0.817 0.664 0.703 0.714 0.815 0.675 0.722 (section with mechanical filter, pump and regulating valve)

Pressure drop on filter in CL of loop 3 YA03DP08 0.729 0.792 0.689 0.731 0.733 0.793 0.706 0.727 Pressure drop on MCP of loop 3 YA03DP09 0.721 0.820 0.655 0.706 0.734 0.817 0.667 0.720 Pressure drop in cold header of SG 3 YB03W01 YA03DP13 0.852 0.855 0.615 0.649 0.853 0.853 0.619 0.649 Pressure drop in hot header of SG 3 YB03W01 YA03DP14 0.922 0.905 0.740 0.717 0.924 0.903 0.754 0.711 Flow rate in the pump circulation loop YD03D01 (I 7) YA03F01 0.653 0.734 0.501 0.410 0.653 0.734 0.501 0.410 Pressure drop in HL of loop 4 (elevation part) YA04DP02 0.765 0.794 0.725 0.731 0.757 0.810 0.718 0.709 Pressure drop in CL of loop 4 YA04DP04 0.852 0.831 0.712 0.759 0.862 0.842 0.712 0.758 (from SG outlet to the flow meter)

Pressure drop in CL of loop 4 YA04DP05 0.856 0.832 0.725 0.747 0.862 0.841 0.726 0.748 (from the flow meter to mechanical filter)

Pressure drop in CL of loop 4 YA04DP06 0.816 0.808 0.615 0.722 0.823 0.819 0.611 0.725 5-10 (section with mechanical filter, pump and regulating valve)

Pressure drop on filter in CL of loop 4 YA04DP08 0.846 0.836 0.686 0.724 0.851 0.841 0.695 0.727 Pressure drop on MCP of loop 4 YA04DP09 0.817 0.805 0.513 0.647 0.824 0.814 0.509 0.651 Pressure drop in cold header of SG 4 YB04W01 YA04DP13 0.834 0.808 0.603 0.705 0.838 0.812 0.605 0.711 Pressure drop in hot header of SG 4 YB04W01 YA04DP14 0.834 0.841 0.754 0.756 0.830 0.838 0.752 0.733 Flow rate in the pump circulation loop YD04D01 (I 8) YA04F01 0.650 0.692 0.525 0.629 0.650 0.693 0.525 0.630 Water level in SG1 YB01L01 0.908 0.886 0.797 0.689 0.902 0.875 0.776 0.692 Pressure of fluid in SG YB01W01 YB01P01 0.954 0.952 0.949 0.892 0.953 0.946 0.941 0.897 Water level in SG2 YB02L01 0.884 0.875 0.786 0.767 0.879 0.871 0.772 0.759 Pressure of fluid in SG YB02W01 YB02P01 0.958 0.937 0.960 0.911 0.958 0.949 0.956 0.917 Water level in SG3 YB03L01 0.927 0.920 0.846 0.837 0.926 0.919 0.833 0.842 Pressure of fluid in SG YB03W01 YB03P01 0.960 0.926 0.965 0.925 0.962 0.942 0.964 0.931 Water level in SG4 YB04L01 0.853 0.884 0.688 0.565 0.854 0.876 0.688 0.565 Pressure of fluid in SG YB04W01 YB04P01 0.955 0.951 0.952 0.897 0.955 0.948 0.945 0.903 Pressure drop on DC (upper part with inlet chamber) YC01DP01 0.842 0.817 0.754 0.760 0.841 0.813 0.756 0.761 Pressure drop on DC (inlet in the middle part) YC01DP02 0.751 0.904 0.584 0.716 0.822 0.911 0.662 0.733 Pressure drop on DC (middle part) YC01DP03 0.722 0.885 0.689 0.731 0.739 0.888 0.683 0.777 Pressure drop on DC (outlet from the middle part) YC01DP04 0.757 0.817 0.776 0.772 0.768 0.817 0.781 0.790 Pressure drop in lower DC chamber YC01DP05 0.815 0.892 0.826 0.830 0.814 0.885 0.828 0.828

Table 12 Evaluation of FOM for CL-4.1-03 Test (Cont'd)

TRACE V5 patch3 TRACE V5 patch4 all W1 W2 W3 all W1 W2 W3 Parameter TAG 0 - 2593 0 - 400 400 - 1470 1470 - 2593 0 - 2593 0 - 400 400 - 1470 1470 - 2593 Pressure drop between DC LC and CS YC01DP06 0.732 0.762 0.750 0.772 0.734 0.788 0.734 0.781 Pressure drop in the channel with FRS (lower part) YC01DP07 0.667 0.753 0.661 0.664 0.658 0.724 0.644 0.660 TRACE V5 patch3 TRACE V5 patch4 all W1 W2 W3 all W1 W2 W3 Parameter TAG 0 - 2593 0 - 400 400 - 1470 1470 - 2593 0 - 2593 0 - 400 400 - 1470 1470 - 2593 Pressure drop in the channel with FRS YC01DP08 0.691 0.756 0.648 0.687 0.689 0.751 0.646 0.679 (inlet in the middle part)

Pressure drop in the channel with FRS YC01DP09 0.674 0.759 0.602 0.672 0.674 0.762 0.602 0.672 (outlet from the middle part)

Pressure drop in the channel with FRS (upper part) YC01DP10 0.753 0.837 0.735 0.705 0.757 0.835 0.731 0.719 Pressure drop between CS and UP YC01DP11 0.691 0.686 0.700 0.678 0.703 0.724 0.707 0.687 Pressure drop in UP (on the insert) YC01DP12 0.710 0.756 0.717 0.702 0.716 0.781 0.732 0.700 Pressure drop in UP (lower part) YC01DP13 0.716 0.725 0.684 0.701 0.737 0.762 0.705 0.726 Pressure drop in UP (middle part) YC01DP14 0.804 0.812 0.684 0.690 0.815 0.831 0.685 0.694 Pressure drop in UP (top piece) YC01DP15 0.797 0.747 0.571 0.439 0.854 0.801 0.609 0.441 Pressure drop between DC and UP YC01DP16 0.714 0.717 0.545 0.674 0.713 0.723 0.535 0.671 5-11 Pressure drop on CS BP YC01DP17 0.723 0.829 0.672 0.681 0.719 0.842 0.637 0.689 Pressure of coolant in UP YC01P16 0.962 0.950 0.855 0.859 0.965 0.955 0.871 0.888 Water level in PRZ YP01L02 0.945 0.942 0.515 0.582 0.944 0.942 0.515 0.582 Pressure of fluid in PRZ YP01B01 YP01P01 0.962 0.948 0.850 0.855 0.966 0.954 0.866 0.896 FOM total avg 0.797 0.802 0.684 0.674 0.804 0.807 0.686 0.681

6 RUN STATISTICS The transients were calculated on calculation server with Intel Xeon E5-2643 v2 processor 3.50 GHz under GNU/Linux CentOS 7.4.1708. The run statistics is shown in the following Table

13. The TRACE patch04 calculation run twice faster than patch03 did due more stable calculation mainly upon transient conditions.

Table 13 Run Statistics TRACE V5.0 patch03 TRACE V5.0 patch04 Number of components 330 330 Number of time steps 249 194 141 812 Transient time 3 800 s 3 800 s CPU time 24 909 s 12 632s CPU time / Transient time 0.153 0.301 6-1

7 CONCLUSIONS The main goal of these analyses was to assess the TRACE code using the cold leg break 4.1%

CL-4.1-03 in the large scale test facility PSB VVER. The second reason was to assess changes of the code TRACE V5.0 from version patch03 to patch04. Before doing the final calculation a complex sensitivity study of the impact of various parameters and code options on the prediction of the rod heaters cladding temperature was performed. For the final calculations the following settings were applied:

• The level tracking model in the core and UP region was activated

• The finer discretization of the upper part of the core (the height of two top layers of the core region is half of the rest of the core)

• TRACE V5.0 patch03 and patch04 were used A part of these analyses is quantitative assessment of agreement of the calculations against the experiment data that can help identify pros and cons of an applied way of modeling integral test facility.

Comparisons of post-test TRACE calculations with experimental data proved that both version of the TRACE code are capable to model PSB-VVER integral system effects reasonably. The calculated time courses of the main facility parameters were similar to that of the test, indicating that all of the significant events that occurred in the test were reproduced by in the calculation To quantify errors/deviations of presented TRACE calculations a set of the figure of merit (FOM) was evaluated using software ACAP. FOMs of 71 main measured and calculated parameters were evaluated analogously for both TRACE calculations. The following table shows the final average FOM evaluated for both calculations at pre-defined time windows of interest.

• W0: 0 ÷ 2593 s - the whole transient

• W1: 0 ÷ 400 s - an early stage of the test when no ECCS flow was provided

• W2: 400 ÷ 1470 s - a middle stage of the test when accumulators flow was provided

• W3: 1470 ÷ 2593 s - the final stage of the test when ECCS LPIS flow was provided Table 14 Final Average FOM Evaluated for Pre-defined Time Windows FOM avg W0 W1 W2 W3 Time 0 - 2593 0 - 400 400 - 1470 1470-2593 0.797 0.802 0.684 0.674 TRACE V5.0 patch03 very good prediction very good prediction good prediction good prediction 0.804 0.807 0.686 0.681 TRACE V5.0 patch04 very good prediction very good prediction good prediction good prediction Presented overall final FOMs (time window W0) prove that the both code versions correctly predicted behavior of test facility during the whole transient, although the version patch 04 prediction seems slightly better. It is clearly visible that better agreements were reached during the early stage of the transient whereas the worst agreements were identified in the end transient. These results correspond to the duration of the CL-4.1-03 test (2593 s), that was 7-1

longer than common LOCA tests with duration of tens of seconds. The accumulation of minor deviations might lead to gradual increasing of deviations of main calculated parameters.

Despite the quantitative assessment gave mainly good or very good predictions of the selected main parameters following particular discrepancies were identified. The TRACE code predicted a little bit slower decrease of liquid water in the upper part of the core during the first stage of the experiment so the first temperature peak between the 97 s and 105 s was not predicted.

Although the calculated water distribution over the reactor vessel and circulation loops was in a good compliance with experiment data, there are two regions where the TRACE code predicted different liquid content: the middle part of the core and loop seals (a part including the MCP and the filter). The calculated liquid content of the middle part of the core was somewhat lower than measured one. The calculated rate of clearance of the loop seals of loops with no break is slower in case the loop#1 and faster in case the loops #2 and #3. But these discrepancies did not significantly affected predicted heat-up of the upper part of the core, where it was expected to happen first.

The calculation cost of the TRACE patch03 calculation was twice higher than TRACE patch04 calculation. The better performance of the TRACE patch04 calculation was influenced by new enhanced features e.g. AUTO backup time step so time coarse of time step was more stable than in case using version patch03 than does not includes these features.

7-2

8 REFERENCES

1. TRACE V5.840 THEORY MANUAL, Field Equations, Solution Methods, and Physical Models; Division of Systems Analysis, Office of Nuclear Regulatory Research, U. S.

Nuclear Regulatory Commission, Washington, DC 20555-0001; March 2013

2. TRACE V5.840 USERS MANUAL, Volume 1: Input Specification; Division of Systems Analysis; Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; February 2014

3. TRACE V5.840 USERS MANUAL, Volume 2: Modeling Guidelines; Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; January 2014

4. TRACE V5.0 Assessment Manual - Main Report, Division of Risk Assessment and Special Projects, Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001

5. TRACE V5.0 Assessment Manual - Addendum 3: V5.0 patch 03, Division of Risk Assessment and Special Projects, Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001

6. TRACE V5.0 Assessment Manual - Addendum 4: V5.0 patch 04, Division of Risk Assessment and Special Projects, Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001

7. TRACE Pressurized Water Reactor Modeling Guidance Division of Risk Assessment and Special Projects, Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; Revision 1; February 2015

8. Staudenmeier J. TRACE Code Modelling Capabilities. Proceedings of TRACE/SNAP User Workshop. Potomac, Maryland, 27 - 29 March 2006

9. OECD Support Group on the VVER Thermal-Hydraulic Code Validation Matrix.

Validation Matrix for the Assessment of Thermal-Hydraulic Codes For VVER LOCA and Transients. NEA/CSNI/R(2001)4. 01 June 2001, Paris, France

10. A. Del Nevo, F.D. Auria, M. Mazini, M. Vylov, I.V. Elkin, A. Suslov; The Design of PSB-VVER Experiments Relevant to Accident Management; Journal of Power and Energy Systems; Vol2, No. 1, 2008

11. O. I. Melikhov, V. I. Melikhov; Analysis of LOCA thermal-hydraulics of VVER 1000.

Description of PSB-VVER Test Facility; Moscow Power Engineering Institute (Technical University), NPP Department; Moscow; 2008

12. O. I. Melikhov, V. I. Melikhov; Analysis of LOCA thermal-hydraulics of VVER 1000.

Single-phase characteristics of the PSB-VVER main circulation pumps; Moscow Power Engineering Institute (Technical University), NPP Department; Moscow; 2008

13. O. I. Melikhov, I. V. Elkin, I.A. Lipatov, A. V. Kapustin, S. M. Nikonov, G. I. Dremin; Pre-test Experiment Proposal Report (Test 3); PSB-VVER Report, PSB-16; OECD NEA; December 2003 8-1

14. O. I. Melikhov, I. V. Elkin, I.A. Lipatov, G. I. Dremin, S. M. Nikonov, S. A. Galchanskaya, A. V. Kapustin,A. A. Rovnov, V. I. Gudkov, A. I. Antonova, A.F. Chalych; Post-test Quick Look Report (Test 3); PSB-VVER Report, PSB-17; OECD NEA; May 2004

15. O. I. Melikhov, I. V. Elkin, I.A. Lipatov, G. I. Dremin, S. M. Nikonov, A. V. Kapustin, A. A.

Rovnov, V. I. Gudkov, A. I. Antonova; Post-test Full Experimental Data Report (Test 3);

PSB-VVER Report, PSB-18; OECD NEA; May 2004

16. V. Blinkov, O.I. Melikhov; Effect of Steam Generator Nodalization on RELAP5 Simulation of Loss-of-Feed-Water Transient in PSB-VVER Facility; Seventh International Information Exchange Forum On SAFETY ANALYSIS FOR NUCLEAR POWER PLANTS OF VVER AND RBMK TYPES; Slovakia, Pieany; 27-31 October 2003

17. P. Heralecký; Engineering Handbook modelu PSB-1.0; TES-Z-09-175; TES s.r.o. Tebí;

prosinec 2009

18. Thermal-Hydraulic Codes News, Volume 09, No.3&4; U.S. Nuclear Regulatory Commission; December 2009

19. Preliminary Safety Assessment Report, Rev.2 - Input Data for Safety Analysis of NPP Dukovany, Revision V5. 01-March-2007

20. WNA Reactor Database. March 3, 2018

21. Petruzzi A. and Auria F. Thermal-Hydraulic System Codes in Nuclear Reactor Safety and Qualification Procedures. Science and Technology of Nuclear Installations Volume 2008, Article ID 460795

22. Matthew D. Lazor; Evaluation of Assessment Techniques for Verification and Validation of the TRACE Nuclear Systems Code; The Pennsylvania State University, College of Engineering; December 2004

23. A. Proek, B. Mavko; Quantitative Code Assessment with Fast Fourier Transformation Based Method Improved by Signal Mirroring; NUREG/IA-0220; Jozev Stefan Institute; Slovenia; Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission; Washington DC, 20555-0001; December 2009 8-2

APPENDIX A INPUT DECK NODALISATION SCHEMES

Figure A-1 TRACE Nodalization Scheme of Primary Circuit of PSB-VVER Facility A-2

Figure A-2 TRACE Nodalization Scheme of Secondary Circuit of PSB-VVER Facility A-3

APPENDIX B THERMAL-HYDRAULIC DIAGRAM AND MEASUREMENT LOCALISATION AT PSB-VVER FACILITY

Figure B-1 PSB-VVER Thermal-Hydraulic Diagram B-2

Figure B-2 PSB-VVER Reactor Model Measurements B-3

Figure B-3 PSB-VVER Loop 1 and SG-1 Model Measurement B-4

APPENDIX C COMPLETE SET OF COMPARISON PLOTS FOR TRACE CALCULATION

Figure C-1 Primary Pressure Figure C-2 Fuel Cladding Temperature (Top of the Core)

C-2

Figure C-3 Pressurizer Level Figure C-4 LPIS Flow (Boundary Condition)

C-3

Figure C-5 Accumulators Levels Figure C-6 Accumulators Pressures C-4

Figure C-7 Fuel Rod Simulator Power (Boundary Condition)

Figure C-8 Core By-pass Power (Boundary Condition)

C-5

Figure C-9 Secondary Side Pressures C-6

Figure C-10 Steam Generators Levels C-7

Figure C-11 Loop 1 Temperatures C-8

Figure C-12 Loop 2 Temperatures C-9

Figure C-13 Loop 3 Temperatures C-10

Figure C-14 Loop 4 Temperatures C-11

Figure C-15 Loop 1 Pressure Differences 1 C-12

Figure C-16 Loop 1 Pressure Differences 2 C-13

Figure C-17 Loop 2 Pressure Differences 1 C-14

Figure C-18 Loop 2 Pressure Differences 2 C-15

Figure C-19 Loop 3 Pressure Differences 1 C-16

Figure C-20 Loop 3 Pressure Differences 2 C-17

Figure C-21 Loop 4 Pressure Differences 1 C-18

Figure C-22 Loop 4 Pressure Differences 2 C-19

Figure C-23 Pressure Differences DP01-DP04 (DC)

C-20

Figure C-24 Pressure Differences DP05-DP08 (Lower Plenum + Lower Part of FRS)

C-21

Figure C-25 Pressure Differences DP09-DP12 (FRS + Lower Part of Upper Plenum)

C-22

Figure C-26 Pressure Differences DP13-DP16 (Upper Part of Upper Plenum)

C-23

Figure C-27 Primary Loops Flow C-24

Figure C-28 Break Flow Figure C-29 Reactor Collapsed Level C-25

NUREG/IA-0502 Post-Test Analysis of Cold Leg Small Break 4.1% at PSB-VVER Facility using TRACE V5.0 March 2019 Petr Heralecky Technical TES Ltd.

Prazska 597 67401 Trebic, Czech Republic Division of Systems Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 K. Tien, NRC Project Manager The U.S. NRC best estimate thermo-hydraulic computer code TRACE V5.0 has been assessed against the 4.1 % cold leg break (CL-4.1-03) experiment at the large-scale test facility PSB-VVER. The PSB-VVER facility is a 1:300 volume-power scaled model of VVER-1000 NPP located in Electrogorsk, Russia. An extensive TRACE input deck of PSB-VVER facility was developed. The model includes all important components of the PSB-VVER facility: reactor, 4 separated loops, pressurizer, HPIS and LPIS ECCS, several break units, main circulation pumps, steam generators, and important parts of secondary circuit. The TRACE (TRAC/RELAP Advanced Computational Engine) is the latest in a series of advanced, best-estimate reactor systems codes developed by the U.S. Nuclear Regulatory Commission in frame of CAMP (Code Application and Maintenance Program). The TRACE code is a component-oriented reactor systems analysis codes designed to analyze light water reactor transients up to the point of significant fuel damage. The original validation of the TRACE code was mainly based on experiments performed on experimental facilities of typical PWR design. There are some different features of VVER design comparing to PWR. Therefore the validation of the thermo-hydraulic codes for VVER types of reactors is often required by national regulators. The presented analysis is the latest in series of TRACE and RELAP5 assessment calculations evaluated at the company TES. The purpose of performed analyses is to extend the validation of the TRACE code focused on VVER type of NPPs and to support applicability of the TRACE code in the Czech Republic.

CAMP (Code Applications and Maintenance Program)

SNAP (Symbolic Nuclear Analysis Program)

Research Centre ez, TSO PSB-VVER test facility OECD PSB project Electrogorsk Research and Engineering Institute (EREC)

Russian Pressurized Water Type Reactor (VVER)

OECD/NEA/CSNI/WGAMA PSB Project.

NUREG/IA-0502 Post-Test Analysis of Cold Leg Small Break 4.1% at PSB-VVER Facility using TRACE V5.0 March 2019