Numerical simulation of loss-of-flow transient in the MYRRHA reactor

Research output: Contribution to report/book/conference proceedingsIn-proceedings paper

Standard

Numerical simulation of loss-of-flow transient in the MYRRHA reactor. / Koloszar, Lila; Planquart, Philippe; Van Tichelen, Katrien; Keijers, Steven.

18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18). ANS - American Nuclear Society, 2019. p. 1-14.

Research output: Contribution to report/book/conference proceedingsIn-proceedings paper

Harvard

Koloszar, L, Planquart, P, Van Tichelen, K & Keijers, S 2019, Numerical simulation of loss-of-flow transient in the MYRRHA reactor. in 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18). ANS - American Nuclear Society, pp. 1-14, 2019 - NURETH18, Portland, United States, 2019-08-18.

APA

Koloszar, L., Planquart, P., Van Tichelen, K., & Keijers, S. (2019). Numerical simulation of loss-of-flow transient in the MYRRHA reactor. In 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (pp. 1-14). ANS - American Nuclear Society.

Vancouver

Koloszar L, Planquart P, Van Tichelen K, Keijers S. Numerical simulation of loss-of-flow transient in the MYRRHA reactor. In 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18). ANS - American Nuclear Society. 2019. p. 1-14

Author

Koloszar, Lila ; Planquart, Philippe ; Van Tichelen, Katrien ; Keijers, Steven. / Numerical simulation of loss-of-flow transient in the MYRRHA reactor. 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18). ANS - American Nuclear Society, 2019. pp. 1-14

Bibtex - Download

@inproceedings{f112c9704d2d4abe828ddb342faab4da,
title = "Numerical simulation of loss-of-flow transient in the MYRRHA reactor",
abstract = "The current paper describes the loss of flow (LOF) transient investigated in the MYRRHA reactor by the means of Computational Fluid Dynamics. This scenario is starting from the nominal operation case then the two pumps stop simultaneously. An unsteady solution with resolved interface was considered with calculating conjugate heat transfer through the relevant structures (with the myrrhaFOAM, OpenFOAM based flow solver [6]). Due to a postulated event (e.g. loss of the electric grid) the pumps are not powered anymore stops. After the detection of the problem (temperature difference above the core rises with 20 degree) the reactor power is stopped by the safety rods (delay of 1 second). The fuel elements, however, continue to generate residual heat according to the decay heat curve. Due to the loss of the pumps, the pressure difference between the cold and the hot plenum is decreasing, which result in a gravitational flow equilibrating the two free surfaces to the same level. The objective of the work was to determine the flow through the core during the coast down of the pumps and eventual flow reversal into the pump/heat-exchanger box due to the gravitational flow. The simulation revealed that after losing power, the LBE flow reverses into the pumps in less than 0.1 seconds according to the simulations. In the core there is a brief moment of reverse flow, too, but only after the core is scrammed, therefore, the loss of cold LBE flow is not causing overheat. Once the core is scrammed, the position of the maximum temperature in the system shifts to the Above Core Structure, where the residual hot plume rising from the core impinges to the Above Core Upper Closure. The levels of the lower and upper plenum equilibrate roughly 20 seconds after the pump failure event.",
keywords = "transient, myrrha reactor, thermohydraulics, RANS",
author = "Lila Koloszar and Philippe Planquart and {Van Tichelen}, Katrien and Steven Keijers",
note = "Score=3",
year = "2019",
month = "8",
day = "18",
language = "English",
pages = "1--14",
booktitle = "18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18)",
publisher = "ANS - American Nuclear Society",
address = "United States",

}

RIS - Download

TY - GEN

T1 - Numerical simulation of loss-of-flow transient in the MYRRHA reactor

AU - Koloszar, Lila

AU - Planquart, Philippe

AU - Van Tichelen, Katrien

AU - Keijers, Steven

N1 - Score=3

PY - 2019/8/18

Y1 - 2019/8/18

N2 - The current paper describes the loss of flow (LOF) transient investigated in the MYRRHA reactor by the means of Computational Fluid Dynamics. This scenario is starting from the nominal operation case then the two pumps stop simultaneously. An unsteady solution with resolved interface was considered with calculating conjugate heat transfer through the relevant structures (with the myrrhaFOAM, OpenFOAM based flow solver [6]). Due to a postulated event (e.g. loss of the electric grid) the pumps are not powered anymore stops. After the detection of the problem (temperature difference above the core rises with 20 degree) the reactor power is stopped by the safety rods (delay of 1 second). The fuel elements, however, continue to generate residual heat according to the decay heat curve. Due to the loss of the pumps, the pressure difference between the cold and the hot plenum is decreasing, which result in a gravitational flow equilibrating the two free surfaces to the same level. The objective of the work was to determine the flow through the core during the coast down of the pumps and eventual flow reversal into the pump/heat-exchanger box due to the gravitational flow. The simulation revealed that after losing power, the LBE flow reverses into the pumps in less than 0.1 seconds according to the simulations. In the core there is a brief moment of reverse flow, too, but only after the core is scrammed, therefore, the loss of cold LBE flow is not causing overheat. Once the core is scrammed, the position of the maximum temperature in the system shifts to the Above Core Structure, where the residual hot plume rising from the core impinges to the Above Core Upper Closure. The levels of the lower and upper plenum equilibrate roughly 20 seconds after the pump failure event.

AB - The current paper describes the loss of flow (LOF) transient investigated in the MYRRHA reactor by the means of Computational Fluid Dynamics. This scenario is starting from the nominal operation case then the two pumps stop simultaneously. An unsteady solution with resolved interface was considered with calculating conjugate heat transfer through the relevant structures (with the myrrhaFOAM, OpenFOAM based flow solver [6]). Due to a postulated event (e.g. loss of the electric grid) the pumps are not powered anymore stops. After the detection of the problem (temperature difference above the core rises with 20 degree) the reactor power is stopped by the safety rods (delay of 1 second). The fuel elements, however, continue to generate residual heat according to the decay heat curve. Due to the loss of the pumps, the pressure difference between the cold and the hot plenum is decreasing, which result in a gravitational flow equilibrating the two free surfaces to the same level. The objective of the work was to determine the flow through the core during the coast down of the pumps and eventual flow reversal into the pump/heat-exchanger box due to the gravitational flow. The simulation revealed that after losing power, the LBE flow reverses into the pumps in less than 0.1 seconds according to the simulations. In the core there is a brief moment of reverse flow, too, but only after the core is scrammed, therefore, the loss of cold LBE flow is not causing overheat. Once the core is scrammed, the position of the maximum temperature in the system shifts to the Above Core Structure, where the residual hot plume rising from the core impinges to the Above Core Upper Closure. The levels of the lower and upper plenum equilibrate roughly 20 seconds after the pump failure event.

KW - transient

KW - myrrha reactor

KW - thermohydraulics

KW - RANS

UR - http://ecm.sckcen.be/OTCS/llisapi.dll/open/34949591

M3 - In-proceedings paper

SP - 1

EP - 14

BT - 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18)

PB - ANS - American Nuclear Society

ER -

ID: 5468728