Lumped parameters analysis of coupled kinetics and thermal-hydraulics for small reactors

Housiadas, C.

doi:10.1016/s0306-4549(01)00107-4

Cited by 43 publications

(8 citation statements)

References 10 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Because of the wide use of MTR as research reactors worldwide, the International Atomic Energy Agency (IAEA) proposed a benchmark setup of a typical MTR core for investigation by many research institutions around the world under both normal and postulated abnormal operating conditions. This benchmark system has been used by most research centers to calibrate their own models and codes, as reported in Housiadas [2,3], Kazeminejad [4], Adorni et al [5], Lu et al [6], and Hamidouche et al [7], to list a few.…”

Section: Introductionmentioning

confidence: 99%

Three-Dimensional, Numerical Investigation of Flow and Heat Transfer in Rectangular Channels Subject to Partial Blockage

Salama

El-Amin

Sun

2014

Heat Transfer Engineering

View full text Add to dashboard Cite

Numerical simulation of flow and heat transfer in two adjacent channels is conducted with one of the channels partially blocked. This system simulates typical channels of a material testing reactor. The blockage is assumed due to the buckling of one of the channel plates inward along its width. The blockage ratio considered in this work is defined as the ratio between the cross-sectional area of the blocked and the unblocked channel. In this work, we consider a blockage ratio of approximately 40%. However, the blockage is different along the width of the channel, ranging from 0% at the end of the channel to 90% in the middle. The channel walls are sandwiching volumetric heat sources that vary spatially as chopped cosine functions. Interesting patterns are highlighted and investigated. The reduction in the flow area of one channel results in the flow redistributing among the two channels according to the changes in their hydraulic conductivities. The results of the numerical simulations show that the maximum wall temperature in the blocked channel is well below the boiling temperature at the operating pressure. INTRODUCTIONThere is no doubt that materials used in the design and operations of nuclear reactors are subjected to a harsh environment that is very unusual. In addition to structural stresses resulting from the interaction of the different components of reactor systems and the thermal stresses associated with the generation of heat within the core, which may be common in several engineering applications, there are also induced stresses that are unique to nuclear reactor cores and may not be found in any other application. For example, the exposure of these material to continuous bombardment with neutrons and fission products (irradiation) results in new phenomena such as swelling of fuel elements and the wear of containment materials. It is therefore important that experimental studies on irradiation effects in fuel and reactor structural materials be conducted to accurately predict their Color versions of one or more of the figures in the article can be found online at www.tandfonline.com/uhte. mechanical behavior and the possible changes in their nuclear chemistry. Unfortunately, it is not possible to mimic such an environment in a single isolated experiment. In other words, the only way to study such effects is within nuclear reactors themselves, which motivated the design of special types of nuclear reactors, called material testing reactors (MTR). These are designed to address in considerable detail the development and the qualification of materials and fuels under irradiation with sizes and environment conditions relevant for nuclear power plants in order to optimize safe operations of existing and future power reactors. MTR typically operate at low temperatures (coolant below 100 • C), the operating conditions are severe in a sense. That is, while power reactor fuel operates at power density of about 5 kW/cm 3 , a research reactor fuel may be at 17 kW/cm 3 in the fuel region. Furthermore, in MTR, bu...

show abstract

Section: Introductionmentioning

confidence: 99%

Three-Dimensional, Numerical Investigation of Flow and Heat Transfer in Rectangular Channels Subject to Partial Blockage

Salama

El-Amin

Sun

2014

Heat Transfer Engineering

View full text Add to dashboard Cite

show abstract

“…For an MTR type fuel element with a rectangular geometry formed by a cooling channel of width W and thickness 2b and a fuel plate of thickness 2d. During the fuel cooling a convective heat transfer is take place between the fuel and the coolant so the local temperatures equations of the coolant and the fuel can be expressed as follow, Housiadas (2002).…”

Section: Lumped Parameter Modelmentioning

confidence: 99%

“…For these reasons, the need of developing simplified and efficient computer codes for MTR research rector uses has received more attention in the last few decades. Firstly, based upon the lumped parameter approach as given by Gaheen et al (2007), Housiadas (2002), Lashkari (2015), and El-Khatib et al (2013) and secondly, based upon the transport equations of fluid flow and heat transfer as given by Lu et al (2009), Bousbia-Salah and Hamidouche (2005) and Al-Yahia et al (2013). In order, to predict more accurately the behavior of the MTR nuclear research reactor during transient operating conditions.…”

Section: Introductionmentioning

confidence: 99%

Lumped parameter model for the safety assessment of the IAEA 10MW benchmark reactor during a protected loss of flow accident

Hamed

2023

Energy Thermofluids Eng.

View full text Add to dashboard Cite

The loss of flow accident in the IAEA 10Mw benchmark reactor is treated in this study through a coupled solution of the point kinetic and thermal hydraulic models. By employing the lumped parameter technic, a one-dimensional thermal hydraulic model is derived and solved beside the point kinetic model with the continuous reactivity feedback effect of the coolant and fuel temperatures. This coupled solution, allows us, to determine, the transient variation of the maximum fuel and coolant temperatures, furthermore, to the safety criteria (ONBR, OFIR, CHFR and BOR) which must be greater than the imposed safety limits, to avoid any nuclear reactor safety issue. Finally, the obtained results of our simulation were validated after a satisfactory comparison with the results of other codes system and developed computer codes.

show abstract

“…PARET code has been used extensively for research reactor analysis which iteratively solves for the neutronic-hydrodynamic-heat transfer aspect of the reactor under steady state and transient behaviour. It can be used to investigate core reactivity insertions being a coupled kinetics and thermal hydraulics code for predicting the course of non-destructive transients in research reactors (Woodruff et al, 1996;Housiadas, 2002;Velit and Primm, 2008).…”

Section: Neutron Kinetic Modelingmentioning

confidence: 99%

Safety Studies and General Simulations of Research Reactors Using Nuclear Codes

Costa¹,

Reis²,

Silva³

et al. 2011

Nuclear Power - System Simulations and Operation

View full text Add to dashboard Cite

Lumped parameters analysis of coupled kinetics and thermal-hydraulics for small reactors

Cited by 43 publications

References 10 publications

Three-Dimensional, Numerical Investigation of Flow and Heat Transfer in Rectangular Channels Subject to Partial Blockage

Three-Dimensional, Numerical Investigation of Flow and Heat Transfer in Rectangular Channels Subject to Partial Blockage

Lumped parameter model for the safety assessment of the IAEA 10MW benchmark reactor during a protected loss of flow accident

Safety Studies and General Simulations of Research Reactors Using Nuclear Codes

Contact Info

Product

Resources

About