Radiation Damage Assessment of Zircaloy and Stainless Steel Cladding materials based on Neutron Flux and Energy Deposition using both Computational Tools and Analytical Solution.

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dc.contributor.advisor Danso, K.A.
dc.contributor.advisor Gyeabuo, A.A.I.
dc.contributor.author Alhassan, S.
dc.contributor.other University of Ghana, College of Basic and Applied Sciences, Department of Nuclear Engineering
dc.date.accessioned 2017-02-02T09:29:13Z
dc.date.accessioned 2017-10-13T17:45:39Z
dc.date.available 2017-02-02T09:29:13Z
dc.date.available 2017-10-13T17:45:39Z
dc.date.issued 2016-07
dc.description Thesis (MPhil) - University of Ghana, 2016
dc.description.abstract The maintenance of the structural integrity of materials in the nuclear reactor is a crucial issue both in-service and out of service. The cladding which forms an integral part of the fuel assembly isolates and prevents the fuel from contaminating the coolant. Pure Zirconium alloys and Steels have extensive use in the nuclear industry including its usage as clad materials for both Pressurized Water Reactor and Boiling Water Reactor fuels. These materials possess good combination of properties such as low neutron absorption, creep behavior, stress-corrosion cracking resistance, reduced hydrogen uptake, reduced corrosion. However, these structural components are susceptible to defects when exposed to high heat, pressure and irradiation. In this regard, the research focused on the use of computational tools to assess the radiation damage on zircaloy and stainless steel clad materials. In this thesis, a modified Low Enriched Uranium core (LEU) input deck was created with focus on the determination of neutron parameters with the MCNP5 code for a typical MNSR operating at 34KW maximum power using Zircaloy-4 fuel clad. The neutron energy deposition and neutron flux (neutron parameters) were employed in the SRIM-TRIM code and analytical radiation model calculations respectively to ascertain the radiation damage. The MCNP5 results as obtained from running the LEU input deck from the Argonne National Laboratory (USA) registered an average energy of 9.871884MeV in all ten lattice rings. The average fast neutron flux representative of all 344 fuel rods gave 5.29667E+11ncm-2s-1 while the average fast neutron flux in the ten lattice rings gave 7.46E+13n/cm2.s. In the SRIM code, the target width was determined as about 2.81μm for neutron interaction and 40.9μm for the radiation interaction. The damage assessment in the TRIM code established Zircaloy-4 as the best cladding material as it recorded the least number of vacancies sustained, least replacement collisions of 147 and recoiling energy of 0.09 whilst Eurofer-97 suffered the highest vacancies created with stainless steel type-308 experiencing the highest replacement collision. The analytical calculations of the radiation damage on Zircaloy-4 using both the Kinchin-Pease and Norgett-Robinson Torrens models was determined as recording displacement of 17 and 14 atoms from the lattice site after 10,000 collisions respectively for only 30 minutes of operation. The calculation of the radiation damage suggests that the zircaloy-4 clad material shows good resistance to defect formation and propagation hence the displacement per atom after a much longer operation time will still leave the clad intact. en_US
dc.format.extent Xvii, 130p. ill.
dc.language.iso en en_US
dc.publisher University of Ghana en_US
dc.title Radiation Damage Assessment of Zircaloy and Stainless Steel Cladding materials based on Neutron Flux and Energy Deposition using both Computational Tools and Analytical Solution. en_US
dc.type Thesis en_US
dc.rights.holder University of Ghana

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