Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core

Adorni, M.; Bousbia-Salah, Anis; Hamidouche, Tewfik; Maro, Beniamino Di; Pierro, F; D'Auria, Francesco Saverio

doi:10.1016/j.anucene.2005.06.001

Cited by 52 publications

(10 citation statements)

References 4 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…In other words, the blockage occurs while the reactor is operating under normal operating conditions. For example, Pierro et al [19] and Adorni et al [5], using the best estimate analysis, simulated blockage scenarios of the IAEA generic MTR using the onedimensional code Relap 5. On the other hand, Lu et al [6] also simulated the same system assuming average channel scenario using also Relap 5 code.…”

Section: Motivationmentioning

confidence: 99%

“…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%

See 1 more Smart Citation

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: Motivationmentioning

confidence: 99%

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

“…Understanding the behavior of fuel elements under abnormal condition is an essential topic in the development of accident scenarios for research reactors [1][2][3][4]. For nuclear power plants the behavior of fuel under accidental conditions is well studied in international projects such as PHEBUS and VERCORS [5][6][7].…”

Section: Introductionmentioning

confidence: 99%

Microstructural analysis of MTR fuel plates damaged by a coolant flow blockage

Leenaers

Joppen

Berghe

2009

Journal of Nuclear Materials

View full text Add to dashboard Cite

“…It may also result in a cladding failure and release of radioactive inventories, which makes it a noteworthy accident to investigate the progression and underlying phenomena. Adorni et al (2005) analyzed the partial and total flow blockage of the standard fuel assembly utilizing RELAP5/Mod3.3 code. They adopted a typical research reactor layout and modeled each 25 fuel assembly (FA) as individual flow paths so that the simulated blockage could represent the assembly-wise transient.…”

Section: Introductionmentioning

confidence: 99%

Transient thermal–hydraulic analysis of complete single channel blockage accident of generic 10 MW research reactor

Son

Yang

Park

et al. 2015

Annals of Nuclear Energy

View full text Add to dashboard Cite

Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core

Cited by 52 publications

References 4 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

Microstructural analysis of MTR fuel plates damaged by a coolant flow blockage

Transient thermal–hydraulic analysis of complete single channel blockage accident of generic 10 MW research reactor

Contact Info

Product

Resources

About