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Analysis for Enhancement of Inherent Safety of an MTR

Muhammad, Farhan (2010) Analysis for Enhancement of Inherent Safety of an MTR. PhD thesis, Pakistan Institute of Engineering & Applied Sciences, Islamabad .

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Abstract

The effects of change in fuel density, change in clad material and change in fuel material on the inherent safety features of a typical material test reactor were analyzed. The International Atomic Energy Agency’s 10 MW benchmark reactor was selected for the study. Standard computer codes WIMS-D4 and CITATION were used to perform neutronics calculations while PARET was used to carryout the steady state and transient thermal hydraulic analysis. In all, seven thermal hydraulic simulations were performed for each configuration. They were the steady state analysis, four controlled transients i.e. fast reactivity insertion, slow reactivity insertion, fast loss of flow and slow loss of flow transients, and two uncontrolled reactivity insertion transients, i.e. small reactivity insertion and large reactivity insertion transients. Two families of the high density dispersion fuels were analyzed to see the effect of changed uranium density on the inherent safety features of the reactor. These families were U3Si2-Al (having uranium densities of 4.10, 4.80 and 5.66 g/cm3) on the lower side and U9Mo-Al (having uranium densities of 6.57, 7.74 and 8.90 g/cm3) on the upper side. It was observed that the steady state thermodynamic behaviour of all the fuels was same, only the fuel temperatures of U3Si2-Al fuels showed some differences. During the fast reactivity insertion transient, the maximum reactor power achieved increased by about 29% for U3Si2 fuel-family while the increase was 45% for U9Mo fuel-family. This resulted in increased maximum temperatures of fuel, clad and coolant outlet, achieved during the transient. This increase for U3Si2 fuels was 32 K, 21.1 K and 5.1 K respectively, while for U9Mo fuels it was 27.7 K, 19.7 K and 7.9 K respectively for maximum fuel, clad and coolant outlet temperatures. During the slow reactivity insertion and loss of flow transients, no appreciable difference in the reactor power and temperature profiles was observed. For small reactivity insertion transient, the new power level increased as uranium density increased. The increase was 8.1% for U3Si2 fuel-family while it was 5.8% for U9Mo fuel-family. In uncontrolled large reactivity insertion transient, the feedback reactivities were unable to control the reactor which resulted in the coolant boiling; the one with the highest fuel density was the first to reach the ONB. In order to see the effects of different fuel materials, the original aluminide (UAlx-Al) fuel of the reactor was replaced with silicide (U3Si-Al and U3Si2-Al) and oxide (U3O8-Al) dispersion fuels having the same uranium density of 4.40 g/cm3 as of the original fuel. The oxide fuel had higher fuel temperatures during steady state and transients. During fast reactivity insertion transient, the maximum power reached for oxide fuel was 0.35 MW lesser than that of aluminide fuel, but its maximum fuel temperature was 13 K higher. With respect to the UAlx-Al fuel, the maximum powers of U3Si-Al and U3Si2-Al fuels were higher by 2.11 MW and 1.82 MW respectively, while the maximum fuel temperatures were lower by 5.7 K and 4.5 K respectively. During slow reactivity insertion and loss of flow transients, the power and temperature profiles of all the fuels were almost the same only fuel temperatures varying; the maximum fuel temperature of the oxide fuel being 8 K to 12 K higher than that of the other fuels. During uncontrolled small reactivity insertion transient, the maximum fuel temperature attained by the oxide fuel was almost 16 K higher than that of the others at the new steady state. During uncontrolled large reactivity insertion transient, the coolant of oxide fuel was the last to reach the ONB but again at the cost of higher fuel temperature. In order to see the effects of different clad materials, only the Al clad and side plates of the reactor fuel were replaced by stainless steel (clad of a fast reactor) and zircaloy-4 (clad of a PWR). The zircaloy-4 clad gave a positive clad temperature feedback coefficient. The very high absorption cross section of stainless steel made it a very unlikely choice for clad material. Out of the remaining two, the main difference was in the fuel temperatures with zircaloy-4 cladded fuel having higher fuel temperatures. The temperature of zircaloy-4 cladded fuel was 20 K to 40 K higher than that of Al cladded fuel during different transients.

Item Type:Thesis (PhD)
Uncontrolled Keywords:Analysis, Enhancement, Inherent, Safety, MTR, fuel, density, clad, material, transients, Uranium, densities
Subjects:Engineering & Technology (e) > Engineering(e1) > Nuclear engineering(e1.24)
ID Code:6367
Deposited By:Mr. Javed Memon
Deposited On:30 Jun 2011 12:22
Last Modified:30 Jun 2011 12:22

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