ANALISIS DOSIS RADIASI PADA JARINGAN TUMOR DENGAN SIMULASI PROGRAM MCNP-5

Direct measurement of each radiation dose to the patient's organs is not possible. In general, to estimate the dose absorbed by human organs is approached by measurements in human phantoms, but this approach is still too rough because the composition of phantoms is not easily made the same as the actual organ composition. Currently, for important matters such as the accuracy of determining the absorption dose by human organs, the Monte Carlo simulation method (MCNP) with special software is used. This has led to a growing desire for scientists to make the transition from using phantoms to computing software for medical physics applications. However, until now no comprehensive document has been written to introduce the use of the MCNP program to simulate its application in medical physics. The purpose of this study was to analyze the absorbed dose of gamma radiation in tumor tissue in the breast by simulating changes in distance and tumor size using the MCNP-5 program. This can be useful in ensuring the application of radiation protection to the patient and the environment in which the patient is located. The results showed that the radiation dose in cell 1 (tumor tissue) with a change in the distance between the radiation source and cell 1 was getting bigger, resulting in a decrease in the dose in cell 1, while the effect of cell volume 1 was greater, the greater the dose received by cell 1. In addition, through this simulation it can be seen that for each addition of 1 cm3 the volume of cell 1 for tumor tissue can increase the absorption energy by 3.5x10e-12 Gray. Keywords: MCNP-5; simulation; radiation dose; tumor tissue AbstrakPengukuran setiap dosis radiasi pada organ pasien tidak dimungkinkan secara langsung. Pada umumnya untuk memperkirakan dosis yang diserap oleh organ tubuh manusia didekati dengan pengukuran pada phantom manusia, namun pendekatan ini juga masih terlalu kasar karena komposisi phantom tidak mudah dibuat sama dengan komposisi organ yang sebenarnya. Sehingga saat ini, untuk hal-hal yang penting seperti ketepatan penentuan dosis serap oleh organ tubuh manusia, digunakan metode simulasi Monte Carlo (MCNP) dengan perangkat lunak khusus. Hal ini mendorong meningkatnya keinginan para ilmuwan melakukan transisi dari penggunaan phantom ke penggunaan komputasi perangkat lunak untuk aplikasi fisika medis. Namun sampai saat ini belum tersedia dokumen komprehensif yang ditulis untuk memperkenalkan penggunaan program MCNP guna mensimulasikan aplikasinya dalam fisika medis. Tujuan penelitian ini adalah menganalisis dosis serap radiasi gamma pada jaringan tumor di payudara melalui simulasi perubahan jarak dan besar tumor menggunakan program MCNP-5. Hal ini dapat berguna dalam memastikan penerapan proteksi radiasi pada pasien dan lingkungan tempat pasien. Hasil penelitian menunjukkan dosis radiasi pada sel 1 (jaringan tumor) dengan perubahan jarak antara sumber radiasi dengan sel 1 semakin besar, mengakibatkan besar dosis di sel 1 semakin menurun, sedangkan pengaruh volume sel 1 yang semakin besar maka dosis yang diterima sel 1 semakin besar juga. Selain itu, melalui simulasi ini dapat diketahui untuk setiap penambahan 1 cm3 volume sel 1 jaringan tumor dapat meningkatkan energi serap sebesar 3,5x10e-12 Gray.

Weakening of neutron and gamma radiation from radioisotope sources by material based on modified titanium hydride

The paper presents experimental studies of the radiation-protective properties of a material based on a modified titanium hydride with respect to gamma and neutron radiation of point radioisotope sources in barrier and continuous protection geometries. The calculated models of the problem of solving the radiation transfer equation for the Monte Carlo method and a comparative assessment of experimental and calculated results is given. The assessment of the amplitude distribution of gamma radiation in the thickness of the material of protection showed a significant reduction in the power of the equivalent dose of radiation gamma in the energy range of 180 – 250 keV, which is due to the effect of the Compton dispersion. The length relaxation of the dose of γ-radiation in 137Сs by the security material was 4.80 ± 0.18 cm. The length of the density relaxation of fast neutrons from the Pu-α-Be source was 6.20 ± 0.18 cm. Comparative analysis of the experimental and calculated data of the protective properties of the material based on modified titanium hydride In relation to radioisotope sources, showed high convergence of the results obtained and the adequacy of the application of the settlement model of the task for the MCNP program used.

PENGARUH DENSITAS URANIUM TERHADAP UMUR BAHAN BAKAR NUKLIR DI DALAM REAKTOR RSG-GAS DITINJAU DARI ASPEK NEUTRONIK

Reaktor Serba Guna G.A. Siwabessy (RSG-GAS) Serpong merupakan reaktor nuklir tipe Material Testing Reactor (MTR). Reaktor ini awalnya dioperasikan menggunakan bahan bakar dispersi U3O8/Al pengkayaan uranium 19,75 % 235U dengan densitas uranium 2,96 gU/cm3. Bahan bakar U3Si2/Al densitas 2,96 gU/cm3 telah berhasil diproduksi dan digunakan sebagai bahan bakar RSG-GAS menggantikan bahan bakar U3O8/Al, sedangkan penelitian bahan bakar berbasis UMo/Al dengandensitas 7 gU/cm3 juga telah diperoleh dalam bentuk pelat mini. Penelitian tentang bahan bakar densitas tinggi masih berfokus pada proses pabrikasi, sedangkan perhitungan tentang umur atau masa pakai (lifetime) dan korelasinya dengan burn up bahan bakar belum banyak dilakukan. Berkaitan dengan hal tersebut, pada penelitian ini dilakukan perhitungan umur bahan bakar dan korelasinya terhadap burn up menggunakan pasangan program ORIGEN dan MCNP. Program ORIGEN digunakan untuk mensimulasikan proses waktu iradiasi, sehingga diperoleh data produk fisi dan uranium sisa (235U tidak mengalami reaksi fisi). Sementara itu, program MCNP digunakan untuk menghitung kritikalitas di dalam teras reaktor. Waktu iradiasi digunakan untuk perhitungan umur bahan bakar, sedangkan kritikalitas digunakan untuk mengetahui burn up maksimal untuk bahan bakar U3Si2/Al dan UMo/Al. Hasil perhitungan menunjukkan bahwa peningkatan densitas uranium berdampak kepada bertambahnya lama iradiasi di dalam reaktor dan burn up bahan bakar. Waktu yang dibutuhkan untuk mencapai burn up 56 % masing masing bahan bakar U3O8/Al; U3Si2/Al, dan U-7Mo/Al selama 188 hari, 292 hari, dan 420 hari. Peningkatan densitas uranium menyebabkan bahan bakar U3O8/Al mampu mencapai burn up 56 %, sedangkan U3Si2/Al dan U-7Mo/Al dapat mencapai nilai burn up sebesar 68,97 % dan 76,76 %. Meningkatnya umur (lifetime) dan burn up bahan bakar berdampak kepada meningkatkan efisiensi bahan bakar di dalam reaktor.

The Effect of MA Transmutation in the PWR on Fuel Cycle

The minor actinides (MA) is important nuclides in the spent fuel which is bad for human ecological environment. Pressurized water reactor (PWR) is the main reactor type at commercial operation around world. It is important to find the appropriate loading patterns when introducing minor actinides to the PWR core. In this paper, we study the effect of MA transmutation in the PWR on fuel cycle. First, we use the MCNP program to simulate the model of PWR and the effective multiplication factor.Then,the MA is introduced into core in different ways and mass to simulate the effective multiplication factor. In conclusion,without considering chemical skim control and control rods, we change the thickness of the MA, until the keff closes to 1, We find that loading minor actinides to burnable poison rods for transmutation is an optimal minor actinide loading pattern.

Thermal Neutron Utilization Factor Calculation by Monte Carlo

The Neutron Transport Theory is accurate in reactor engineering analysis, but the calculation process is tedious and complicated. The objective of the present study obtains the thermal utilization factor f by Monte Carlo method. The study establishes the pressurized water reactor model by MCNP program firstly, and calculates the dioxide pellets, zirconium alloy cladding and moderator’s neutron flux density distribution. The thermal neutron disadvantage factor ζ will be gained according to the definition formula. Based on the functional relationship between the above thermal neutron disadvantage factor ζ and the thermal utilization factor f, this study finally obtains the thermal utilization factor f.