PRELIMINARY RESULTS OF THE SEVENTH THREE-DIMENSIONAL AER DYNAMIC BENCHMARK PROBLEM CALCULATION.

SOLUTION WITH DYN3D AND RELAP5-3D CODES.

Marek Benćík, Jan Hádek

Nuclear Research Institute ež plc, 250 68 ež, Czech Republic E-mail: [email protected]

ABSTRACT

The paper gives a brief survey of the 7th three-dimensional AER dynamic benchmark ¬ calculation results received with the codes DYN3D and RELAP5-3D at NRI ež. This benchmark was defined at the 20th AER Symposium in Hanassari (Finland). It is focused on investigation of transient behaviour in a VVER-440 nuclear power plant. Its initiating event is opening of the main isolation valve and re-connection of the loop with its main circulation pump in operation. The VVER-440 plant is at the end of the first fuel cycle and in hot full power conditions. Stationary and burnup calculations were performed with the code DYN3D. ¬ Transient calculation was made with the system code RELAP5-3D . The two-group homogenized cross sections library HELGD05 created by HELIOS code was used for the generation of reactor core neutronic parameters. The detailed six loops model of NPP th ¬ Dukovany was adopted for the 7 AER dynamic benchmark purposes. The RELAP5-3D full core neutronic model was coupled with 49 core thermal-hydraulic channels and 8 reflector channels connected with the three-dimensional model of the reactor vessel. The detailed nodalization of reactor downcomer, lower and upper plenum was used. Mixing in lower and upper plenum was simulated. The first part of paper contains a brief characteristic of ¬ RELAP5-3D system code and a short description of NPP input deck and reactor core model. The second part shows the time dependencies of important global and local parameters.

1. INTRODUCTION

The seventh three-dimensional hexagonal dynamic AER benchmark problem was defined as a re-connection of the loop No.1 with its main circulation pump in operation in a VVER- 440/213 plant [1]. The reactor core is in hot full power conditions at the end of its first full cycle. The basic reactor core geometric and material parameters were given. No nuclear cross section data as well as kinetic data were available in the benchmark definition. It means that one’s own best estimate neutronic data have to be used in the calculation. The benchmark participants have to use also their own plant input data decks developed according to the needs of their own NPP system codes. Only main geometric parameters of primary and secondary circuit and main characteristics of control and safety systems were given in order to bring into agreement the input data decks. The calculations of the seventh dynamic AER benchmark problem were performed by using of codes DYN3D [2] and RELAP5-3D [3] at the Nuclear Research Institute (NRI, ÚJV) ež.

The reason for NRI ež participation at AER benchmark project is the verification of RELAP5-3D system code solution (especially testing of reactor vessel model) against the results received by other best estimate three-dimensional neutronic codes coupled with NPP system or CFD codes – APROS (VTT Espoo, Finland) [4], DYN3D/ATHLET, ANSYS CFX/DYN3D (HZDR, Germany) [5] and other.

2. RELAP5-3D SYSTEM CODE

RELAP5-3D is the latest in the series of RELAP5 codes [3]. The RELAP5 (Reactor Excursion and Leak Analysis Program) series of codes has been developed at the Idaho National Engineering and Environmental Laboratory under the U. S. Nuclear Regulatory Commission, the U. S. Department of Energy, and a consortium of several countries and domestic organizations sponsorship.

RELAP5 – at NRI ež has been tested on data from experimental facilities - PMK-NVH (Hungary), PACTEL (Finland), RVS-3 (NRI ež – Czech Republic), and on experimental data from transients and tests performed at NPP Dukovany. The process of RELAP5 testing goes on permanently at NRI ež.

The RELAP5-3D version contains several important enhancements over previous versions of the code. The most important feature that distinguishes the RELAP5-3D code from the previous versions is the fully integrated, multidimensional thermal-hydraulic and neutron kinetic modelling capacity.

The multi-dimensional component in RELAP5-3D was developed to allow the user to more accurately model the multi-dimensional flow behaviour that can be exhibited in any component or region of a LWR system. Typically this will be the reactor downcomer, lower plenum, core, upper plenum and upper head regions. However, the model is general, and is not restricted to use in the reactor vessel. For example, it can be used for describing of steam generator. The component defines a one, two, or three-dimensional array of volumes and the internal junctions connecting them.

The geometry can be either Cartesian (x, y, z) or cylindrical (r, θ, z). The number of volumes in a three-dimensional component is limited to 999 volumes.

The multi-dimensional neutron kinetic model in RELAP5-3D is based on the NESTLE code [3], [6] which solves the two or four group neutron diffusion equations in either Cartesian or hexagonal geometry.

All thermal-hydraulics is removed from the original NESTLE code in its RELAP5-3D version and instead of it the RELAP5-3D thermal-hydraulic model is used. It means that so called internal coupling between RELAP5-3D thermal-hydraulics and NESTLE kinetic model exists.

3. INPUT MODEL OF NPP WITH VVER-440/213

RELAP5-3D was tested on the input deck of Nuclear Power Plant (NPP) Dukovany with VVER-440/213 reactor (the 1st unit).

The input model of NPP Dukovany used for the analyses is a detailed 6-loop model which was modified according to the 7th AER dynamic benchmark specifications. The input deck originates in an “international” 3-loops input deck of NPP Jaslovské Bohunice prepared under IAEA regional program “Safety Assessment of VVER-440/213” [7]. This model has been further enlarged and modified according to the specification of NPP Dukovany and needs for more detailed analyses. The current version of the input model is characteristic with 6-loop configuration, detailed modelling of ECCS systems and detailed modelling of both main steam system and feed water system on the secondary side [8].

3-D TH reactor pressure vessel model has been validated on available transient data from NPP Dukovany (e.g. asymmetrical loop cooling) [9].

The nodalization scheme of the input model is presented in Fig.1 (reactor vessel) and Fig. 2 (primary circuit).

3.1 Main characteristics of input model

Main characteristics of nodalization: ° 6 modelled loops, ° RPV consists of 1 three-dimensional (MULTID) object: - 8 azimuthal sectors, - 4 radial sectors, - 17 axial sectors, - some volumes (related to core and reactor head) are disabled (by setting flow areas = 0), - normal one-dimensional equations are used on each of the coordinate direction (MULTID → Word 7 = 1 on Card CCC0001), ° core (including reflector) is simulated by bundle of 1-D channels (pipes) connected to the multidimensional lower plenum and upper plenum, ° detailed steam generator (SG) nodalization, ° detailed description of steam lines (MSL) and main steam headers (MSH), ° detailed description of feed water (FW) system.

Main modelled operational control systems: ° reactor power control system (ARM-5S), ° reactor power limiting system (ROM), ° pressurizer (PRZ) heaters and spraying, ° primary make-up (TK) and let-down (TE), ° turbine control system (TVER), ° steam bypass valves to main condenser (BRU-C).

Main modelled safety systems: ° reactor trip signals AZ-1, 2, 3, 4, ° Engineered Safety Features Actuation System (ESFAS) signals, ° all Technological Protection of Steam Generator (TPSG) signals, ° all Local Protection of Steam Generator (LPSG) signals, ° safety valves at PRZ and SGs, ° atmospheric dump valves (BRU-A), ° Emergency Core Cooling System (ECCS), ° High Pressure Injection System (HPIS), ° Low Pressure Injection System (LPIS), ° Accumulators (HA).

3.2 Nodalization of VVER/440-213 core

The reactor core configuration is shown in Fig. 3. The core was divided into 12 axial layers (2 layers for unheated lower and upper part, 10 layers for heated core part with total length of 244 cm). The both unheated parts were treated as axial reflectors. The location of control rod groups is shown in Fig. 4. The core was surrounded with a hypothetical radial band of reflector cassettes. 349 fuel assemblies are assigned to 49 core thermal channels (pipes). Reflector cassettes are represented by 8 reflector channels (connected to the 3rd radial sector of MULTID object). Every fuel channel consists of 12 axial volumes (1 non-heated volume – lower reflector, 10 active core volumes, and 1 non-heated volume – upper reflector). The details of reactor core thermal-hydraulic model and its coupling with 3-D neutron kinetic model are shown in Fig. 5.

3.3 Neutron cross sections model

RELAP5-3D USER option was selected for the calculations [3]. In this option the user creates his own external subroutine which computes the set of cross sections for a single node given by the material type in the node determined from the composition maps, the region average properties in the node specified in the zone maps for node, and the control fractions and insertion directions for any control rods associated with the node. The user may also specify the value of up to four additional variables in each node that may affect the neutron cross sections (fuel burnup, Xe and Sm concentration etc.). This subroutine is further linked ¬ with the RELAP5-3D code. The exchange of parameters between user procedure (subroutine userxs.f) and RELAP5-3D reactor core model can be seen in Fig. 6.

3.4 Macroscopic cross sections library

The two-group macroscopic cross section library CSLIBR (see Fig. 6) for homogenized fuel assemblies, control rods and reflectors as well as feedback coefficients and reactor kinetics parameters can be generated from HELIOS neutronic library [10], with using of subprogram DYLIE [11]. The VVER-440 control rod model consists usually from three to four parts - fuel follower, connecting node (without or with Hf plates), and absorber part. In our RELAP5- 3D calculations the control rod model consisted only from fuel follower and absorber part was considered. The volumetric portions of moderator and steel in reflector and control rod assemblies were obtained from the technologic documentation in order to generate corresponding neutronic cross sections from neutronic library.

4. THE SEVENTH THREE-DIMENSIONAL AER DYNAMIC BENCHMARK PROBLEM

This benchmark was defined at the 20th AER Symposium in Hanassari (Finland) [1]. It concerns a re-connection of an isolated circulation loop with lower temperature in a VVER- 440/213 plant. The core is in full power conditions (1196.25 MW) at the end of its first fuel cycle. The control rods belonging to the 6th group of control rods are at position of 175 cm from the bottom of the core. Other groups of control rods are fully withdrawn. The initial state conditions of the core at the beginning of the transients were given. The gas-gap heat transfer coefficient is to be kept constant during the whole transient: 3000 W/m2K. Radial thermal conductivities and thermal capacities of fuel pellet and cladding should be described with best available data of each participant. The main geometrical parameters of the plant system and the characteristics of control and safety systems to be considered were given.

The initiating event is a re-connection of a loop No. 1 with coolant temperature of 100 oC. The scenario of loop connection is following [1]: First opens the MIV in the cold leg, after that the MCP is started. After the MCP reached full flow the MIV in the cold leg starts to open. A water slug with reduced temperature enters the core that results in increase of reactor fission power. The activation of the reactor scram is caused by the corresponding power level signal (110% of the nominal power, nominal power = 1375 MW) with the delay of 0.5 s. The stuck rod belonging to control rod group No. 4 (position 293) is located in the sector with highest overcooling (see Fig. 4). As a result of the scram, the turbines are turned-off by closing the turbine isolation valves. All MCPs remain in operation. After SCRAM feed water temperature decreases linearly from 220°C to 160°C (within 50 s).

5. PROGRAM TOOLS FOR BENCHMARK SOLUTION

According to the 7th AER Dynamic Benchmark definition [1] the solution consists of both initial stationary and burnup reactor core calculation and transient calculation.

Stationary and burnup calculations were performed with the code DYN3D [2]. Transient ¬ calculation was made with the system code RELAP5-3D [3].

6. STATIONARY AND BURNUP CALCULATIONS

The benchmark transient was suggested for the end of the first fuel cycle, therefore the burnup calculation for the fresh loading of the VVER-440 reactor core (Fig. 3) was necessary. This calculation was performed at the nominal power level of 1375 MW with fully withdrawn and unchanged positions of control rods. The results of initial stationary and burnup calculations are given in Table 1 and Fig. 7.

Table 1: Results of initial stationary and burnup calculations

Teff [FPD] cb [g H3BO3 / kg H2O] keff 0 6.965 0.999989 365 0.0099 1.000001

7. RESULTS OF TRANSIENT CALCULATION

Transient was initiated at the time t = 0.0 s by reconnection of isolated loop No 1. At first hot MIV in the leg opened, after that MCP was started. When MCP reached full flow, MIV in the cold leg began to open (at the time 20.0 s). Slug of colder water reached the core that resulted in increase of neutron power due to kinetics feedbacks. At the time 33.8 s reactor power level reached 110% of the nominal level, SCRAM was activated and control rods began to drop. Maximum core power (1585.8 MW) was reached shortly after reactor SCRAM at the time 34.5 s. Reconnection of loop No 1 caused decrease of core inlet temperatures not only by slug of colder water contained in this loop but also by heat transfer in its SG (due to lower temperature in SG 1). Distinct asymmetry in core inlet temperatures remained till the temperatures in SGs equalized (at the end of scenario at the time 500 s). Turbine isolation valve closing started at the time 43.8 s (10 s after reactor SCRAM with delay 0.5 s). Decrease of primary temperatures led to the decrease of pressurizer level and consequently to the start of make up pumps (first pump at the time 36.1 s and second pump at the time 76.1s)

The initial thermohydraulics conditions, core powers and system events reached during the transient are given in Tables 2 to 3. Selected snapshots of 3-D fission power distributions are presented in Figures 8 to 12. Figures 13 to 20 show behaviour of important time functions.

Table 2: Initial termohydraulic conditions

Proposed Calculated Basic parameters: Unit conditions conditions

Primary circuit Reactor power MW 1196.25 1196.25 Upper plenum pressure MPa 12.26 12.26 Core inlet temperature oC 267.4 267.6

Proposed Calculated Basic parameters: Unit conditions conditions

Loop No 1 cold leg temperature °C 100.0 101.0 (isolated loop) Loop mass flow rate kg/s 1470.0 1464.6-1469.8 Core bypass mass flow rate % 3 3 Pressurizer collapsed level m 5.97 5.97 Secondary circuit Pressure at SG outlet MPa 4.63 4.61-4.67 Feed water temperature oC 220.0 220.0 Feed water mass flow rate kg/s 124.5 127.6-130.3 SG collapsed level m 2.015 2.015

Table 3: Sequence of main events

Time [s] Event -500.0 Beginning of calculation 0.0 MIV of the hot leg No. 1 starts to open 2.0 MCP No. 1 starts 20.0 MIV of the cold leg No. 1 starts to open 33.8 SCRAM – power level of the 110 % of the nominal level 34.5 Maximum core power (=1585.8 MW) 36.1 1st TK pump starts 43.8 Closing of turbine isolation valves starts 76.1 2nd TK pump starts 500.0 End of calculation

8. CONCLUSIONS

The preliminary results of the seventh three-dimensional dynamic AER benchmark problem calculations are presented in this paper. The stationary and burnup calculations were performed with the code DYN3D. The transient calculation was made with the system code RELAP5-3D. The two-group homogenized cross section library HELGD05 created by the HELIOS code was used for the generation of best estimate neutronic input data. The standard NRI ež RELAP5-3D input deck for VVER-440/213 was adopted according to the benchmark proposal.

Several topics have arisen from the calculation of such complex AER dynamic benchmark problem for the RELAP5-3D current and future application:

° Accurate RELAP5-3D tuning of initial steady state. The relatively accurate estimate of initial distribution of coolant temperatures, energy releasing, heat structure temperatures, and coolant mass flow rates in coolant channels is necessary in order to reach the proper stabilization of steady-state before the transient calculation.

° The multidimensional thermal-hydraulic model of the reactor vessel was included into system input deck of the VVER-440/213. 8 azimuthal sector nodalization of reactor vessel was prepared. Mixing in lower and upper plenum was simulated.

° The multi-channels reactor core representation (49 channels model) was prepared. Possible influence of multi-channel geometry on safety important criteria such as for example the maximum fuel pellet centerline temperature should be studied.

10. NOMENCLATURE

1-D one-dimensional 3-D three-dimensional ARM-5S reactor power control system AZ-1,2,3,4 reactor trip signals BRU-A atmospheric dump valves BRU-C steam bypass valves to main condenser cb boron acid concentration CFD computational fluid dynamics CL cold leg CRs control rods DC downcomer ECCS Emergency Core Cooling System ESFAS Engineered Safety Features Actuation System FAV fast acting valve FPD full power day FW feed water HA hydro-accumulators HL hot leg HP hermetic volume HPIS High Pressure Injection System IAEA International Atomic Energy Agency keff effective coefficient of multiplication LP lower plenum LPIS Low Pressure Injection System LPSG Local Protection of Steam Generator LWR light water reactor MCP main circulation pump MIV main isolation valve MSH main steam header MSL main steam line MSLB main steam line break MULTID RELAP5-3D multidimensional component NPP Nuclear Power Plant NRI, ÚJV Nuclear Research Institute (ež) RPV reactor pressure vessel PRZ pressurizer

ROM reactor power limiting system SCRAM Safety Control Rod Ax Man SG steam generator TE let-down system Teff effective time TH thermohydraulic THX0 X-division of Low Pressure Injection System TH1X hydro-accumulator TJX0 X-division of High Pressure Injection System TK make-up system TPSG Technological Protection of Steam Generator Tm moderator temperature TVER turbine control system UP, HSK upper plenum ¬ USER option in RELAP5-3D code VVER Russian type of pressurized water reactor

11. REFERENCES

[1] Kotsarev, A., Lizorkin, M. and Petrin, M.: Definition of the 7th Dynamic AER Benchmark – VVER-440 Pressure Vessel Coolant Mixing by Re-connection of an Isolated Loop (Edition 1), Proceedings of the 20th Symposium of AER on VVER Reactor Physics and Reactor Safety, Hanasaari, Espoo, Finland, September 2010, pp. 583-598, KFKI Atomic Energy Research Institute, Budapest (2010), ISBN 978-963- 372-644-0

[2] Grundmann, U., Mittag, S., Rohde U. and Kliem, S.: DYN3D Version 3.2 (FORTRAN90) Code for Calculation of Transients in Light Water Reactors (LWR) with Hexagonal or Quadratic Fuel Elements. Code Manual and Input data Description for Release, Forschungszentrum Dresden-Rossendorf e.V., Germany, December 2009

[3] RELAP5-3D© Code Manual, Volumes 1, 2, 3, 4, 5. Idaho National Laboratory, Idaho Falls, Idaho 83415, INEEL-EXT- 98-00834, Revision 2.4, June 2005

[4] Syrjalähti, E.: Preliminary Calculation of 7th AER Benchmark, AER Working Group D Meeting on VVER Safety Analysis, KTH Stockholm, Sweden, April 2011

[5] Kliem, S., Grahn, A, Kiss, B. and Kozmenkov, Y: Status of Simulation of the 7th AER Benchmark with DYN3D, ATHLET and ANSYS CFX, AER Working Group D Meeting on VVER Safety Analysis, KTH Stockholm, Sweden, April 2011

[6] NESTLE 5.0 – Code System to Solve the Few-Group Neutron Diffusion Equation Fixed-Source Steady- State and Transient Problems, RSIC, P.O. Box 2008, Oak Ridge, TN 37831-6362, December 1996

[7] Engineering Handbook for the Safety Analysis of WWER-440 Model 213 Nuclear Power Plants. Revision 1. IAEA TC/RER/004-A042, April 1994

[8] Král, P., Krhounková, J., Denk, L. and Parduba, Z.: Thermal Hydraulic Analyses of LOCA with Break Size D50 for PTS Evaluation, ÚJV ež, November 1999 (in Czech)

[9] Benćík, M., Hádek, J.: Development of RELAP5-3D model for VVER-440 Reactor,, 2010 RELAP5 International User’s Seminar, West Yellowstone, Montana, September, 2010

[10] Tinková, E., Tinka, I.: Dukovany NPP, Safety Analysis EDU Support, HELGD05 Library Description, Report ÚJV ež – ENERGOPROJEKT PRAHA division, EGP 5014-F-050519, Praha, November 2005 (in Czech)

[11] Hádek, J.: Description of the Code Dylie for Automatized Preparation of the DYN3D Code Neutronic Input Data, Report ÚJV ež, ÚJV 11051 R, T, December 1997 (in Czech)

Pressure vessel RELAP5 model

17

B - B

16

7 6 CL 1 15 CL 6

8 14 5 HL CL 5 3 4 13 1 2

UP 11 CL 2 B B 10 1 4 CL 9 CL4 CL 3 2 8 3

7 HA HA

A 6 A

CORE A - A

5 7 6 DC 8 5

4 1 2 3 4

1 4

3

2 3

2 LP 1

Fig. 1 Reactor vessel nodalization TH13 TH12 TH11 TH10 333 RV

331 PV 1

330 332 PV 2 338 350 344 336 335 SM 343 345 349 YP11S01/S02 317 348 316 05 TE SYSTEM 346 04 (TK) 342 329 347 03 CL 1/2 CL 5/3 CL 6/4 (124/154) (254/184) (284/224) 10 10 PT TF10 07 07 356 02 293 295 297 340 01 355 (S15) 351 06 06 HP (S01) 328 290 TY11 292 294 296

358 354 352 02 01 291 357 353 (S17) (S02) 327 05 05 TK SYSTEM CL 1/2 CL 5/3 CL 6/4 105 (122/152) (252/182) (282/222) 326 PRESSURIZER 194 197 299 430 420 410 400 324 SPRAY 04 04 190 192 195 198

323 193 196 199 03 03 191 322

321 02 02

MSH 707 01 01 325

705 REACTOR 311 312 HEATERS STEAM GENERATOR 1 706 108 118 432 422 412 402 03 FW 17 09 703 310 EFW 433 423 413 403 08 07 314 07 07 434 424 414 404 315 109 119 07 306 702 07 701 02 06 06 117 06 405 06 01 02 03 04 05 115 06 07 08 09 10 06 435 425 415 16 305 05 05 01 02 03 04 05 116 06 07 08 09 10 05 04 05 05 303 406 03 304 04 04 04 01 02 03 04 05 114 06 07 08 09 10 04 04 436 426 416 320 02 03 03 03 01 02 03 04 05 113 06 07 08 09 10 03 03 427 407 04 01 437 02 02 01 02 03 04 05 112 06 07 08 09 10 02 438 418 417 15 03 302 02 02 01 02 03 04 05 06 07 08 09 10 428 02 301 01 01 01 111 01 01 319 110 408 14 300 UPPER PLENUM UPPER PLENUM 101 700 DOWNCOMER DOWNCOMER HL 1 03 120 13 100 102 01 02 MIV 02 11 127 125 03 04 05 104 106 128 02 126 124 105 02 10 123 103 478 MIV MCP 9 03 CL 1 121 477 02 490 02 122 05 04 03 492 8 476 496 7 470 493 497 472 374 384 394 494 498 6 HL 4 (202-05) HP HP 375 385 395 473 454 452 CL 4 453 474 (226-01) CL 2 CL 3 CL 5 5 458 463 483 456 450 (156) (186) (256) 457 HP HP HP HP

468 488 443 373 383 393 4 462 482 469 466 489 486 467 487

461 481 460 480 448 378 388 398 TH60 TH20 3 442 372 382 392 449 446 379 376 389 386 399 396 447 377 387 397 2 441 371 381 391 440 370 380 390 TH40 TJ20 TJ40 TJ60 1

Fig. 2 Primary circuit nodalization

Fig. 3 Distribution of fuel assemblies in VVER-440 reactor core

Fig. 4 Location of control rod groups in VVER-440 reactor core

346 347 348 349 85 86 337 338 339 340 341 342 343 344 345

325 326 327 328 329 330 331 332 333 334 335 336 76 95 312 313 314 315 316 317 318 319 320 321 322 323 324 83 84 296 297 298 299 300 301 302 303 304 305 306 307 308 309 310 311 349 fuel assemblies 75 74 93 96 279 280 281 282 283 284 285 286 287 288 289 290 291 292 293 294 295 82 261 262 263 264 265 266 267 268 269 270 271 272 273 274 275 276 277 278 73 94 242 243 244 245 246 247 248 249 250 251 252 253 254 255 256 257 258 259 260 72 81 92 224 225 226 227 228 229 230 231 232 233 234 235 236 237 238 239 240 241 66 71 91 205 206 207 208 209 210 211 212 213 214 215 216 217 218 219 220 221 222 223 64 24 26 185 186 187 188 189 190 191 192 193 194 195 196 197 198 199 200 201 202 203 204 21 49 core channels (1D pipes) 166 167 168 169 170 171 172 173 174 175 176 177 178 179 180 181 182 183 184 62 61 22 23 146 147 148 149 150 151 152 153 154 155 156 157 158 159 160 161 162 163 164 165 65 63 25 127 128 129 130 131 132 133 134 135 136 137 138 139 140 141 142 143 144 145 8 reflector channels (1D pipes connected to 51 31 27 rd 109 110 111 112 113 114 115 116 117 118 119 120 121 122 123 124 125 126 the 3 radial sector of MULTID object) 52 41 32 90 91 92 93 94 95 96 97 98 99 100 101 102 103 104 105 106 107 108 54 33 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 42 35 Every fuel channel has 12 axial volumes: 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 56 53 34 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 • 1 non-heated volume - lower reflector 44 43 26 27 28 29 30 31 32 33 34 35 36 37 38 • 10 active core volumes 55 36 14 15 16 17 18 19 20 21 22 23 24 25 • 1 non-heated volume – upper reflector 5 6 7 8 9 10 11 12 13 46 45 1 2 3 4

Fuel

Reflector

175 Fuel assembly number 21 RELAP5-3D channel (pipe) Fig. 5 TH core model and coupling with 3-D NK

Fig. 6 Neutronic data preparation and exchange of parameters in USER option

8

boron acid concentration [g H3BO3/ kg H2O] 6.965 7 ] O

2 6 H g k / 3 O

B 5 3 H g [ n o i t 4 a r t n e c n o c

3 d i c a n o r o 2 b

1

365 0 0 50 100 150 200 250 300 350 400 cycle operation days [EFPD]

Fig. 7 Boron acid concentration

1.4 1.2 1.4 1 1.2 0.8 0.6 1 0.4 Loop 6 Loop 1 0.8 0.2 0.6 0

0.4

0.2

0

Loop 5

Loop 2

Total power = 1196.25 MW, Fission power = 1117.1 MW

Loop 4 Loop 3

0 0.2 0.4 0.6 0.8 1 1.2 1.4

Fig. 8 Normalized fission power at 0 s

1.6 1.4 1.6 1.2 1.4 1 0.8 1.2 0.6 1 0.4

0.8 0.2 0 0.6

0.4 Loop 6 Loop 1

0.2

0

Loop 5

Loop 2

Total power = 1415.1 MW, Fission power = 1335.8 MW

Loop 4 Loop 3

0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6

Fig. 9 Normalized fission power at 32.5 s

1.8 1.6 1.8 1.4 1.2 1.6 1 1.4 0.8 1.2 0.6 0.4 1 0.2 0.8 0 0.6

0.4 Loop 6 Loop 1 0.2 0

Loop 5 Control rods start to drop Loop 2

Total power = 1585.8 MW, Fission power = 1504.1 MW

Loop 4 Loop 3

0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8

Fig. 10 Normalized fission power at 34.5 s

2.5

2 2.5 1.5 2 1

1.5 0.5

0 1

0.5 Loop 6 Loop 1

0

Loop 5

Loop 2 Group K6 fully inserted

Total power = 376.5 MW, Fission power = 303.1 MW

Loop 4 Loop 3

0 0.5 1 1.5 2 2.5 Fig. 11 Normalized fission power at 40 s

4 3.5 4 3 3.5 2.5 2 3 1.5 2.5 1

2 0.5 0 1.5 Loop 6 Loop 1 1

0.5

0

Loop 5

Loop 2

Total power = 49.9 MW, Fission power = 2.24 MW Loop 4 Loop 3

0 0.5 1 1.5 2 2.5 3 3.5 4

Fig. 12 Normalized fission power at 150 s

1,26E+07

1,24E+07

1,22E+07

1,20E+07

1,18E+07

1,16E+07 ] a P [

1,14E+07 e r u s s 1,12E+07 e r P

1,10E+07

1,08E+07

1,06E+07

1,04E+07

1,02E+07

1,00E+07 -100 0 100 200 300 400 500 Time [s] Fig. 13 Primary pressure – upper plenum

1600

1400

1200 ]

s 1000 / g k [ e t a r 800 w o l f s s a 600 M

400

200

0 -100 0 100 200 300 400 500 Time [s]

Loop No. 1 Loop No. 2 Loop No. 3 Loop No. 4 Loop No. 5 Loop No. 6 Fig. 14 Loop mass flow rate

7

6

5

4 ] m [

l e v e L 3

2

1

0 -100 0 100 200 300 400 500 Time [s] Fig. 15 Pressurizer collapsed level (from the bottom)

1,6E+09

1,4E+09

1,2E+09

1,0E+09 ] W [

r 8,0E+08 e w o P

6,0E+08

4,0E+08

2,0E+08

0,0E+00 -100 0 100 200 300 400 500 Time [s]

Fig. 16 Core power

5,0E+08

4,0E+08

3,0E+08 ] W [

r 2,0E+08 e w o P

1,0E+08

0,0E+00

-1,0E+08 -100 0 100 200 300 400 500 Time [s] SG1 SG2 SG3 SG4 SG5 SG6 Fig. 17 Power transferred to SG

350

300

250 ] C [ 200 e r u t a r e p m e 150 T

100

50

0 -100 0 100 200 300 400 500 Time [s] CL1 CL2 CL3 CL4 CL5 CL6 Tsat Fig. 18 Cold leg temperatures – loop seal

350

300

250

300 ] C [

200 e

r 275 u t ] a r C [

e e r p u t

m 150 a 250 r e e T p m e T

100 225

200 50 20 30 40 50 60 70 80 90 100 Time [s]

0 -100 0 100 200 300 400 500 Time [s]

HL 1 HL 2 HL 3 HL 4 HL 5 HL 6

Fig. 19 Hot leg temperatures – loop seal

270

260

250 ] C ° [ e r u t 240 a r e p m e T

230

220 Channels 91-96 are related to loop 1

210 -100 0 100 200 300 400 500 Time [s]

TH channel 21 TH channel 22 TH channel 23 TH channel 24 TH channel 25 TH channel 26 TH channel 27 TH channel 31 TH channel 32 TH channel 33 TH channel 34 TH channel 35 TH channel 36 TH channel 41 TH channel 42 TH channel 43 TH channel 44 TH channel 45 TH channel 46 TH channel 51 TH channel 52 TH channel 53 TH channel 54 TH channel 55 TH channel 56 TH channel 61 TH channel 62 TH channel 63 TH channel 64 TH channel 64 TH channel 66 TH channel 71 TH channel 72 TH channel 73 TH channel 74 TH channel 75 TH channel 76 TH channel 81 TH channel 82 TH channel 83 TH channel 84 TH channel 85 TH channel 86 TH channel 91 TH channel 92 TH channel 93 TH channel 94 TH channel 95 TH channel 96

Fig. 20 Core inlet temperatures