FACTORS AFFECTING INFANT MORTALITY RATE IN KARANGASEM, BALI

The purpose of this study was to determine the factors that influence the infant mortality rate in Karangasem, Bali. The method used in this research is the Log Linier model. In the Log linear model analyze relationship pattern among group of categorical variables which include an association of two or more variables, either simultaneously or partially. A Patterned relationship between variables can be seen from the interaction between variables. Log linear analysis does not distinguish between explanatory variables and response variables. The population in this study was all babies born in Karangasem in 2015 that is as many as 7,895 babies with live birth status and as many as 7,835 babies and 60 infants died. As a sample, 100 babies were taken, of which 60 were live and 40 died. The results show that infant mortality is affected by infant weight, how old the mother during childbirth, and interaction between birth spacing and infant weight

Neutron Characterization of BNCT Water Phantom Based on Kartini Research Reactor Using PHITS

Boron Neutron Capture Therapy (BNCT) is a therapy that utilizes the interaction of thermal neutrons with a boron-10 core that produces alpha particles and lithium nuclei. The result of this boron reaction has high linear energy transfer (LET). BNCT has an advantage over other radiation therapy in that it has a high selectivity level. This research was run using PHITS simulation to find out the value of neutron flux spread over a water phantom. The conclusion of the research is the distribution of neutron flux in the water phantom without boron is higher the the distribution of neutron flux in the water phantom containing boron

Distribution of Water Phantom BNCT Cyclotron based Using PHITS

This research purpose is to estimate the dose distribution of BNCT in water phantom. Some common methods in the treatment of cancer such as brakhiterapi, surgery, chemotherapy, and radiotherapy still have the risk of damaging healthy tissue around cancer cells. BNCT is a selectively-designed technique by targeting high-loaded LET particles to tumors at the cellular level. BNCT proves to be a powerful method of killing cancer without damaging normal tissue. The source of the neutron used from the cyclotron dose in water phantom with the size of 30 cm x 30 cm x 30 cm was calculated using PHITS program. The result from the simulation is that boron water panthom has a dosimetri higher than phantom water without boron.

Neutron Chareacterization of BNCT Water Phantom Based on 30 MeV Cyclotron Using PHITS Computational Code

Cancer is the second leading cause of death globally and was responsible for 8,8 million deaths in 2015. Approximately 70% of deaths from cancer occur in low- and middle-income countries. The war on cancer has been fought with three tools – surgery (cut), radiation therapy (burn) including radiotherapy and bracytherapy, and also chemotherapy (poison). Cancer therapy has increased life expectancy of patients but each treatment modality has its own effects, complications and toxicity. Moreover we have found a new effective method to fight cancer, that is, Boron Neutron Capture Therapy (BNCT). Boron Neutron Capture Therapy (BNCT) has for many decades been advocated as an innovative form of radiotherapy that, in principle, has the potential to be the ideal form of treatment for many types of cancers. This research’s aim is the characterising neutron of BNCT water phantom based on 30 MeV cyclotron using PHITS computational code. The result from the simulation is that thickness of the water phantom, related to flux neutron.

Dose Analysis in Boron Neutron-capture Cancer Therapy (BNCT) Neutron Generator Based for Breast Cancer

The purpose of this study is to know the concentration of boron and irradiation times which optimizes the treatment of breast cancer using the BNCT method. This research was conducted by using MCNPX simulation which outputs are flux neutron, neutron scattering dose and gamma dose. The neutron source used is the BSA D-D Neutron generator model. The independent variable of this research is the boron concentration injected into the cancer. The dependent variable is the total dose rate and irradiation time which determines the effectiveness of BNCT therapy. The controlled variables are the output of the neutron flux, dose and gamma neutron scattering dose. The results showed that in the range of 70-150 µg/g, the dose rate received by cancer increases with increasing the concentration of boron-10. If the dose rate is increased, the irradiation time interval will be faster. The Boron dose of 70 μg/g and the dose rate of irradiation 0.00293603 Gy/sec needs an irradiation time of 409.43 minutes; the boron dose of 90 µg/g and the dose rate of irradiation 0.00241049 Gy/sec needs an irradiation time of 345.71 minutes; the boron dose of 110 µg/g and the dose rate of irradiation 0.00271236 Gy/sec needs an irradiation time of 307.24 minutes; the boron dose of 130 µg/g and the dose rate of irradiation 0.00303389 Gy/sec needs an irradiation time of 274.67 minutes; and the boron dose 150 µg/g and the dose rate of irradiation 0.00334565 Gy/sec needs an irradiation time of 249.08 minutes. The Optimum concentration of boron is 150 µg/g with irradiation time of 249.08 minutes.

In Vivo Total Dose Analysis in Mice for BNCT Trial TRIGA Kartini Research Reactor Based Using PHITS

Boron neutron capture therapy (BNCT) is an effective radiotherapy modality to kill cancer. BNCT can selectively kill cancer cells without damaging the healthy tissue around it by using alpha particle and lithium ion from the reaction of 10B(n,α)7Li. These particles have a track of more or less 5 to 9 μm which is the same as the cell diameter. In order to support the development of BNCT in Indonesia an in vivo simulation is performed in a simple mouse geometry containing a 4T1 breast cancer characteristic treated with BNCT using PHITS program. The Neutron source that was used in this simulation was based on TRIGA Kartini Research Reactor. The boron compound concentration in the tumor was varied from 20 ppm up to 90 ppm, and then the total dose was calculated in the mice. Total dose that the tumor received was 0.0161, 0.0168, 0.0175, 0.0182, 0.0185, 0.0188, 0.019, and 0.0191 Gy-Eq/s, respectively and the irradiation time to reach 50 Gy was 51, 50, 48, 46, 45, 45, 44, 44, 40 minutes respectively. This shows that the higher the concentration of boron compound in the tumor the higher the dose that mice received and irradiation time was decreased with the increase of the boron compound concentration.

MEASUREMENT OF YTTRIUM-90 BIODISTRIBUTION IN SELECTIVE INTERNAL RADIATION THERAPY (SIRT)

Measurement of radionuclides biodistribution in post-radioembolization 90Y SIRT is a part of treatment evaluation, in which the assessment of biodistribution is used to evaluate the possible extrahepatic presence and the absorbed dose estimation for the tumor cells, healthy liver cells, and critical organs. As the dose-response analysis is performed based on this evaluation, the biodistribution measurement coming from post-imaging modality has a crucial role in achieving these goals. The two devices, Single Photon Emission Tomography (SPECT) and Positron Emission Tomography are discussed in some aspects, including the quality of quantitative images, performance characteristics, and absorbed dose considerations.

DESIGN AND MANUFACTURING DEVICE MOBILITY LOAD MOTION SHIELDING PARAFFIN RADIATION PROTECTION SYSTEM FACILITY TEST IN VITRO IN VIVO

Practical work at PSTA-BATAN to find paraffin design and the design of mobile devices with Monte Carlo N Particle (MCNP) software. The method used is to determine the paraffin design and calculate the volume of paraffin. The resulting intact writing that modeled with the MCNP. Shielding is required to absorb the leaking radiation until the 20 mSv / year Dose Limit Value for radiation workers is met. The material used is paraffin. Calculation is done by using MCNPX calculation facility with tariff of 10,42 μSv / hour. The paraffin design criteria are built on recommendations from Indonesian Journal of Physics and Nuclear Applications Volume 1, Number 1, February 2016. Some of the above-standard methods are overcome with the protection aspects of distance and radiation time. Paraffin used is made of hydrocarbons suitable for strengthening shielding structures and for absorbing gamma radiation.

In Vitro and In Vivo Test of Boron Delivery Agent for BNCT

BNCT is an alternate therapy for treating cancer. The principle of BNCT involves a neutron boron uptake and a fission reaction that produce alpha particles and Li ions with a high level of linear energy transfer in the tissue. It is effective in killing tumor cells. To administer boron in the tumor cells, a boron delivery agent is needed. Thus far, there are a variety of boron delivery agents that have been developed. To date, just two main boron-based drugs, BPA and BSH, have been used for clinical studies. Many other boron delivery agents have been evaluated in vivo and in vitro but have not been evaluated clinically. Therefore, the other boron delivery agents have not been used in BNCT clinical studies.

MONTE CARLO N PARTICLE EXTENDED (MCNPX) RADIATION SHIELD MODELLING ON BORON NEUTRON CAPTURE THERAPY FACILITY USING D-D NEUTRON GENERATOR 2.4 MeV

Based Studies were carried out to analyze the internal dose of radiation for workers at Boron Neutron Capture Therapy (BNCT) facility base on Cyclotron 30 MeV with BSA and a room that was actually designed before. This internal dose analyzation included interaction between neutrons and air. The air contained N2 (72%), O2 (20%), Ar (0.93%), CO2, Neon, Kripton, Xenon, Helium and Methane. That internal dose to the worker should be below the dose limit for radiation workers which is an amount of 20 mSv/years. From the particles that are present in the air, only Nitrogen and Argon can change into radioactive element. Nitrogen-14 activated to Carbon-14, Nitrogen-15 activated to Nitrogen-16, and Argon-40 activated to Argon-41. Calculation using tally facility in Monte Carlo N Particle version Extended (MCNPX) program for calculated Neutron flux in the air 3.16x107 Neutron/cm2s. The room design in the cancer facility has a measurement of 200 cm in length, 200 cm in width, and 166.40 cm in height. Neutron flux can be used to calculate the reaction rate which is 80.1x10-2 reaction/cm3s for carbon-14 and 8.75x10-5 reaction/cm3s. The internal dose exposed to the radiation worker is 9.08E-9 µSv.