**2. Methodology**

todetermine the timedevelopment ofmaterial composition andradionuclide inventoryduring theentireoperationoftheplant.Thedeterminationofplutonium(Pu)inenvironmental samples is also important for investigating radiotoxicity and the long-term radiation effects on hu‐ mans that are attributed to Pu rather than the other transuranic elements [1]. The determina‐ tion of the elemental and isotopic compositions of Pu in spent nuclear fuels is required for fuel characterization [2], and also the management of radioactive waste [3]. According to the U.S Nuclear Regulatory Commission, the elements contained in high-level waste include plutoni‐ um-239 (half-life: 24,000 years) and plutonium-240 (half-life: 6800 years). The content of Pu isotopes in high burnup spent nuclear fuel samples (33.21–59.03 GWd/MtU) were determined by alpha and mass spectrometry [4].While it may be necessary to determine Pu contentin spent fuel due to the extremely long half-life of Pu, it is also important to quantify spent fuel based on the total concentrations of key actinides and non-actinides to estimate the entire radioactive

Much research work has been done in the area of sustainable nuclear energy. Some research work on nuclear burnup calculations focused on the analysis of the accuracy of the Monte Carlo Nuclear Particle (MCNP) Code as compared to other computer codes such as BUCCAL 1, GETERA Code and others. Accuracy of the new Monte Carlo based burnup computer code has been studied [5]. The research involved direct usage of neutron absorption tally generated by MCNP5 to determine new nuclide inventories. This eliminates the need to calculate neutron fluxes as input to other computer codes to generate the nuclide inventories and thus incorpo‐ rates fully the capabilities of MCNP. Code-to-code comparison of BUCAL 1 with several burnup codes showed good agreement. The research shows the accuracy of the Monte Carlo Code in predicting the depletion and generation of radionuclide inventories. Monte Carlo burnup simulations of isotope correlation experiments have also been determined using the NPP Obrigheim [6]. In the experiment, the Monte Carlo Nuclear Particle eXtended (MCNPX) code was found to be accurate in predicting the uranium fuel depletion, the plutonium production and the buildup of most of the fission products in a nuclear reactor. The direct integration of the cinder.dat file as the decay code for MCNPX makes it more accurate than in previous experiments when the MCNP transport code is simply linked to other codes such as

The GETERA code [7] was used to calculate fuel burnup in a miniature neutron source reactor (MNSR). A list of 19 radionuclides, including 13 fission products and 6 actinides, was identified and the total radioactivity of the MNSR at the end of the reactor core life was calculated based on these nuclides. This was found to be 9.462 × 1013 Bq. This research used highly enriched uranium as fuel in the MNSR core and does not provide comparisons of the radiological hazards associated with other fuels. The results from the research indicated that the amount of 235U and 238U after burnup decreased about 13.669 and 0.065 g, respectively, thus pre‐

The next section gives details on the Monte Carlo methods for simulating the fuel burnup in nuclear power plants in this research. The three different fuel grades, mixed oxide fuel (MOX), uranium oxide fuel (UOX) and commercially enriched uranium (CEU), would be used for the

dicting again the accuracy of the code in estimating the depletion of uranium.

hazard posed by these fuels.

38 Nuclear Material Performance

ORIGEN.
