READY special report 1

Application of TRAB-3D to BWR and PWR transient calculations

Antti Daavittila, Anitta Hämäläinen, Riitta Kyrki-Rajamäki

Research output: Chapter in Book/Report/Conference proceedingChapter or book articleProfessional

Abstract

TRAB-3D is the newest member in the reactor dynamics code family developed in VTT. It is now possible to analyze both PWRs and BWRs taking into account the real core loading. The core geometry can be either rectangular, as in most western reactors, or hexagonal, as in e.g. VVER reactors. In addition to the core models the codes include models for the cooling circuits and reactor control systems. This paper describes the neutronics/thermal-hydraulics coupling methodology and validation of the three-dimensional TRAB-3D for the application to PWR and BWR transient calculations. Code development in the fields of reactor physics and dynamics, as well as in thermal-hydraulics has been one of the key areas of reactor safety research in Finland since the middle of the seventies. In reactor dynamics, which combines neutron kinetics, heat conduction, heat transfer and hydraulics of the cooling circuit, the first product was a LWR core model (Rajamäki 1980a), which is basically one-dimensional two-group code, but includes a synthesis model for radially nonuniform dynamics. The same core model was applied in the development of the BWR dynamics code TRAB (Rajamäki 1980b and Räty et al. 1991), which is continuously used in the safety analysis of the Finnish TVO power plant. It models the core and the main circulation system inside the reactor vessel, including the steam dome with related systems, steam lines, recirculati-on pumps, incoming and outgoing flows, as well as control and protection systems. Typical applications are the pump trip and the steam line isolation. The code has also been used for the simulation of the RBMK type reactor in accident conditions. The HEXTRAN (Kyrki-Rajamäki 1995) core dynamics model has been developed on the basis of the stationary two-group diffusion code HEXBU-3D (Kaloinen et al. 1981) for hexagonal geometries, with the fuel heat conduction and channel hydraulics description of the TRAB code. TRAB-3D (Kaloinen & Kyrki-Rajamäki 1997) is a similar code used for rectangular lattice reactors. In addition to the core dynamics TRAB-3D also includes the BWR circuit models of TRAB. The validation work of TRAB-3D consists of the successful calculation of OECD benchmarks, the most recent of which has been the PWR main steam line break (MSLB) benchmark (Ivanov et al. 1999), as well as comparisons with measurements of real TVO plant transients (Daavittila et al. 2000). As the cooling circuit model of PWRs serves the SMABRE code (Miettinen 1985) coupled with HEXTRAN or TRAB-3D. SMABRE is a fast running simulation code and it was developed for such thermal-hydraulic accidents as small breaks, originally in order to facilitate parameter variations to support slow RELAP calculations. Its range of application was, however, soon extended to cover such types of accidents as small break LOCA, primary-secondary leak, steam line break, ATWS and most parts of a large break LOCA in VVER plants.
Original languageEnglish
Title of host publicationFINNUS: The Finnish Research Programme on Nuclear Power Plant Safety
Subtitle of host publicationInterim Report 1999 - August 2000
Place of PublicationEspoo
PublisherVTT Technical Research Centre of Finland
Pages150-160
ISBN (Electronic)951-38-5751-4
ISBN (Print)951-38-5750-7
Publication statusPublished - 2000
MoE publication typeD2 Article in professional manuals or guides or professional information systems or text book material

Publication series

NameVTT Tiedotteita - Research Notes
PublisherVTT
Number2057
ISSN (Print)1235-0605
ISSN (Electronic)1455-0865

Fingerprint

Steam piping systems
Hydraulics
Accidents
Loss of coolant accidents
Networks (circuits)
Cooling
Heat conduction
Pumps
Hydraulic equipment
Geometry
Domes
Flow control
Dynamic models
Power plants
Neutrons
Steam
Physics
Heat transfer
Control systems
Kinetics

Cite this

Daavittila, A., Hämäläinen, A., & Kyrki-Rajamäki, R. (2000). READY special report 1: Application of TRAB-3D to BWR and PWR transient calculations. In FINNUS: The Finnish Research Programme on Nuclear Power Plant Safety: Interim Report 1999 - August 2000 (pp. 150-160). Espoo: VTT Technical Research Centre of Finland. VTT Tiedotteita - Research Notes, No. 2057
Daavittila, Antti ; Hämäläinen, Anitta ; Kyrki-Rajamäki, Riitta. / READY special report 1 : Application of TRAB-3D to BWR and PWR transient calculations. FINNUS: The Finnish Research Programme on Nuclear Power Plant Safety: Interim Report 1999 - August 2000. Espoo : VTT Technical Research Centre of Finland, 2000. pp. 150-160 (VTT Tiedotteita - Research Notes; No. 2057).
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abstract = "TRAB-3D is the newest member in the reactor dynamics code family developed in VTT. It is now possible to analyze both PWRs and BWRs taking into account the real core loading. The core geometry can be either rectangular, as in most western reactors, or hexagonal, as in e.g. VVER reactors. In addition to the core models the codes include models for the cooling circuits and reactor control systems. This paper describes the neutronics/thermal-hydraulics coupling methodology and validation of the three-dimensional TRAB-3D for the application to PWR and BWR transient calculations. Code development in the fields of reactor physics and dynamics, as well as in thermal-hydraulics has been one of the key areas of reactor safety research in Finland since the middle of the seventies. In reactor dynamics, which combines neutron kinetics, heat conduction, heat transfer and hydraulics of the cooling circuit, the first product was a LWR core model (Rajam{\"a}ki 1980a), which is basically one-dimensional two-group code, but includes a synthesis model for radially nonuniform dynamics. The same core model was applied in the development of the BWR dynamics code TRAB (Rajam{\"a}ki 1980b and R{\"a}ty et al. 1991), which is continuously used in the safety analysis of the Finnish TVO power plant. It models the core and the main circulation system inside the reactor vessel, including the steam dome with related systems, steam lines, recirculati-on pumps, incoming and outgoing flows, as well as control and protection systems. Typical applications are the pump trip and the steam line isolation. The code has also been used for the simulation of the RBMK type reactor in accident conditions. The HEXTRAN (Kyrki-Rajam{\"a}ki 1995) core dynamics model has been developed on the basis of the stationary two-group diffusion code HEXBU-3D (Kaloinen et al. 1981) for hexagonal geometries, with the fuel heat conduction and channel hydraulics description of the TRAB code. TRAB-3D (Kaloinen & Kyrki-Rajam{\"a}ki 1997) is a similar code used for rectangular lattice reactors. In addition to the core dynamics TRAB-3D also includes the BWR circuit models of TRAB. The validation work of TRAB-3D consists of the successful calculation of OECD benchmarks, the most recent of which has been the PWR main steam line break (MSLB) benchmark (Ivanov et al. 1999), as well as comparisons with measurements of real TVO plant transients (Daavittila et al. 2000). As the cooling circuit model of PWRs serves the SMABRE code (Miettinen 1985) coupled with HEXTRAN or TRAB-3D. SMABRE is a fast running simulation code and it was developed for such thermal-hydraulic accidents as small breaks, originally in order to facilitate parameter variations to support slow RELAP calculations. Its range of application was, however, soon extended to cover such types of accidents as small break LOCA, primary-secondary leak, steam line break, ATWS and most parts of a large break LOCA in VVER plants.",
author = "Antti Daavittila and Anitta H{\"a}m{\"a}l{\"a}inen and Riitta Kyrki-Rajam{\"a}ki",
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Daavittila, A, Hämäläinen, A & Kyrki-Rajamäki, R 2000, READY special report 1: Application of TRAB-3D to BWR and PWR transient calculations. in FINNUS: The Finnish Research Programme on Nuclear Power Plant Safety: Interim Report 1999 - August 2000. VTT Technical Research Centre of Finland, Espoo, VTT Tiedotteita - Research Notes, no. 2057, pp. 150-160.

READY special report 1 : Application of TRAB-3D to BWR and PWR transient calculations. / Daavittila, Antti; Hämäläinen, Anitta; Kyrki-Rajamäki, Riitta.

FINNUS: The Finnish Research Programme on Nuclear Power Plant Safety: Interim Report 1999 - August 2000. Espoo : VTT Technical Research Centre of Finland, 2000. p. 150-160 (VTT Tiedotteita - Research Notes; No. 2057).

Research output: Chapter in Book/Report/Conference proceedingChapter or book articleProfessional

TY - CHAP

T1 - READY special report 1

T2 - Application of TRAB-3D to BWR and PWR transient calculations

AU - Daavittila, Antti

AU - Hämäläinen, Anitta

AU - Kyrki-Rajamäki, Riitta

PY - 2000

Y1 - 2000

N2 - TRAB-3D is the newest member in the reactor dynamics code family developed in VTT. It is now possible to analyze both PWRs and BWRs taking into account the real core loading. The core geometry can be either rectangular, as in most western reactors, or hexagonal, as in e.g. VVER reactors. In addition to the core models the codes include models for the cooling circuits and reactor control systems. This paper describes the neutronics/thermal-hydraulics coupling methodology and validation of the three-dimensional TRAB-3D for the application to PWR and BWR transient calculations. Code development in the fields of reactor physics and dynamics, as well as in thermal-hydraulics has been one of the key areas of reactor safety research in Finland since the middle of the seventies. In reactor dynamics, which combines neutron kinetics, heat conduction, heat transfer and hydraulics of the cooling circuit, the first product was a LWR core model (Rajamäki 1980a), which is basically one-dimensional two-group code, but includes a synthesis model for radially nonuniform dynamics. The same core model was applied in the development of the BWR dynamics code TRAB (Rajamäki 1980b and Räty et al. 1991), which is continuously used in the safety analysis of the Finnish TVO power plant. It models the core and the main circulation system inside the reactor vessel, including the steam dome with related systems, steam lines, recirculati-on pumps, incoming and outgoing flows, as well as control and protection systems. Typical applications are the pump trip and the steam line isolation. The code has also been used for the simulation of the RBMK type reactor in accident conditions. The HEXTRAN (Kyrki-Rajamäki 1995) core dynamics model has been developed on the basis of the stationary two-group diffusion code HEXBU-3D (Kaloinen et al. 1981) for hexagonal geometries, with the fuel heat conduction and channel hydraulics description of the TRAB code. TRAB-3D (Kaloinen & Kyrki-Rajamäki 1997) is a similar code used for rectangular lattice reactors. In addition to the core dynamics TRAB-3D also includes the BWR circuit models of TRAB. The validation work of TRAB-3D consists of the successful calculation of OECD benchmarks, the most recent of which has been the PWR main steam line break (MSLB) benchmark (Ivanov et al. 1999), as well as comparisons with measurements of real TVO plant transients (Daavittila et al. 2000). As the cooling circuit model of PWRs serves the SMABRE code (Miettinen 1985) coupled with HEXTRAN or TRAB-3D. SMABRE is a fast running simulation code and it was developed for such thermal-hydraulic accidents as small breaks, originally in order to facilitate parameter variations to support slow RELAP calculations. Its range of application was, however, soon extended to cover such types of accidents as small break LOCA, primary-secondary leak, steam line break, ATWS and most parts of a large break LOCA in VVER plants.

AB - TRAB-3D is the newest member in the reactor dynamics code family developed in VTT. It is now possible to analyze both PWRs and BWRs taking into account the real core loading. The core geometry can be either rectangular, as in most western reactors, or hexagonal, as in e.g. VVER reactors. In addition to the core models the codes include models for the cooling circuits and reactor control systems. This paper describes the neutronics/thermal-hydraulics coupling methodology and validation of the three-dimensional TRAB-3D for the application to PWR and BWR transient calculations. Code development in the fields of reactor physics and dynamics, as well as in thermal-hydraulics has been one of the key areas of reactor safety research in Finland since the middle of the seventies. In reactor dynamics, which combines neutron kinetics, heat conduction, heat transfer and hydraulics of the cooling circuit, the first product was a LWR core model (Rajamäki 1980a), which is basically one-dimensional two-group code, but includes a synthesis model for radially nonuniform dynamics. The same core model was applied in the development of the BWR dynamics code TRAB (Rajamäki 1980b and Räty et al. 1991), which is continuously used in the safety analysis of the Finnish TVO power plant. It models the core and the main circulation system inside the reactor vessel, including the steam dome with related systems, steam lines, recirculati-on pumps, incoming and outgoing flows, as well as control and protection systems. Typical applications are the pump trip and the steam line isolation. The code has also been used for the simulation of the RBMK type reactor in accident conditions. The HEXTRAN (Kyrki-Rajamäki 1995) core dynamics model has been developed on the basis of the stationary two-group diffusion code HEXBU-3D (Kaloinen et al. 1981) for hexagonal geometries, with the fuel heat conduction and channel hydraulics description of the TRAB code. TRAB-3D (Kaloinen & Kyrki-Rajamäki 1997) is a similar code used for rectangular lattice reactors. In addition to the core dynamics TRAB-3D also includes the BWR circuit models of TRAB. The validation work of TRAB-3D consists of the successful calculation of OECD benchmarks, the most recent of which has been the PWR main steam line break (MSLB) benchmark (Ivanov et al. 1999), as well as comparisons with measurements of real TVO plant transients (Daavittila et al. 2000). As the cooling circuit model of PWRs serves the SMABRE code (Miettinen 1985) coupled with HEXTRAN or TRAB-3D. SMABRE is a fast running simulation code and it was developed for such thermal-hydraulic accidents as small breaks, originally in order to facilitate parameter variations to support slow RELAP calculations. Its range of application was, however, soon extended to cover such types of accidents as small break LOCA, primary-secondary leak, steam line break, ATWS and most parts of a large break LOCA in VVER plants.

M3 - Chapter or book article

SN - 951-38-5750-7

T3 - VTT Tiedotteita - Research Notes

SP - 150

EP - 160

BT - FINNUS: The Finnish Research Programme on Nuclear Power Plant Safety

PB - VTT Technical Research Centre of Finland

CY - Espoo

ER -

Daavittila A, Hämäläinen A, Kyrki-Rajamäki R. READY special report 1: Application of TRAB-3D to BWR and PWR transient calculations. In FINNUS: The Finnish Research Programme on Nuclear Power Plant Safety: Interim Report 1999 - August 2000. Espoo: VTT Technical Research Centre of Finland. 2000. p. 150-160. (VTT Tiedotteita - Research Notes; No. 2057).