Pemanfaatan Energi Nuklir Sebagai Sumber Daya Wahana Angkasa Luar

Kata kunci: pemanfaatan energi nuklir, sumber daya wahan

Penyelesaian analitik persamaan diferensial parsial simultan linier dar\u27 proses transport suatu zat yang mengalami reaksi bertingkat order satu pada medium terbuka dengan sumber titik seketika (Instantaneous Point Sourc

ABSTRACS Transport process of a substance that undergoes sequential first order reaction in open medium occurs in engineering and environmental analysis. In environmental analysis, this problem occurs M the problem of transport of chemicals and radioactive contaminans which undergo reaction process that produces other substances that still have contamination hazard properties. These substances disperse to the environtnient with their own transport properties. Thus the overall contamination analysis must deal with simultaneous partial linier differential equation of each substances. Numerical solution of this problem suffers the limitation of computer memory and the difficulty of boundary condition formulation because.of the fact that environmental medium must be treat as an open medium. Analytical solution can be obtained with succesive general Fourier and Laplace tranform. The solution is still in integral forms and must be solved by numerical integration. However numerical integration does not need very large computer memory space compared with direct numerical solutions. The method has been succesfully demonstrated to calculate concentration of radioactive contaminan and its daughter that undergo sequential decay process in a river stream as time and position function. Key word: penyelesaian analitik, distribusi radio aktif, diferensial parsial linier simulta

Desain Reaktor Nuklir Bermoderator Air Berat Berpendingin Uap Panas Lanjut yang Memiliki Sifat Keselamatan Melekat

Abstract The problem of positive power feedback in the recent PHWR-CANDU design will be overcome by the use of "dual moderator concept ", in which two moderator systems are used, i.e. a main moderator outside the calandria tube and an annular moderator inside the annular space. The numerical calculations have been performed for two core design namely HWR-DMST and HWR-DM-X1 which can reach burn up of 16000 and 17500 MWd / ton U respectively. The results for the two designs is that the values of k at fully anular moderator filling condition are 1.0054 (HWR-DM-ST) and 1.0019 (HWR-DM-X1), while at completelly empty annular moderator condition are 0.9634 (HWR-DM-ST) and 0.9143 (HWR-DM-X1). The decreasing of coolant flow rate from 3043 kg/s to 853 kg/s gives a decreasing of k values of O.0109 (HWRDM- ST) and 0.0232 (HWR-DM-X1). The increasing of inlet coolant enthalpifrom 2950 kJ/kg to 3175 kJ/kg gives a decreasing ofk values of 0.0074 (HWR-DM-ST) and 0.0239 (HWR-DM-X1). Thus it can be summarized that these HWR-DM designs have negative power reactivity feedback. These designs can achieve a thermal efficiencies of 38,5 % with the fuel utilizations of 120,75 and 110,52 kg of natural uranium / kWe-year for HWR-DM-ST and HWR-DM-X1 respectively. Keywords: PHWR-CANDU, dual moderator concep

Conceptual Design of Collimator at Boron Neutron Capture Therapy Facility with 30 MeV Cyclotron and Target 9Be as Neutron Generator Using Monte Carlo N-Particle Extended Simulator

The optimization of collimator has been studied which resulted epithermal neutron beam for Boron Neutron Capture Therapy (BNCT) using Monte Carlo N Particle Extended (MCNPX). Cyclotron 30 MeV and 9Be target is used as a neutron generator. The design criteria were based on recommendation from IAEA. Mcnpx calculations indicated by using 25 cm and 40 cm thickness of PbF2 as reflector and back reflector, 15 cm thickness of TiF3 as first moderator, 35 cm thickness of AlF3 as second moderator, 25 cm thickness of 60Ni as neutron filter, 2 cm thickness of Bi as gamma filter, and aperture with 20 cm of diameter size, an epithermal neutron beam with an intensity  1.21 × 109 n.cm-2.s-1, fast neutron and gamma doses per epithermal neutron of 7.04 × 10-13  Gy.cm2.n-1 and 1.61 × 10-13 Gy.cm2.n-1, minimum thermal neutron per epithermal neutron ratio of 0.043, and maximum directionality of 0.58, respectively could be produced. The results have not passed all the IAEA’s criteria in fast neutron component and directionality

Analysis of Radiation Effects on Workers and Environment Pilot Plant Boron Neutron Capture Therapy (BNCT)

BNCT is a new method in nuclear technology. The aim of BNCT application is to reduce human risk which used to kills cell targeting characteristic. The impact of using this technology should be considered before it is applied, among the effects of radiation on workers and the surrounding environment BNCT pilot plant. A research on modeling of BNCT pilot plant used a collimator for a 30 MeV cyclotron neutron sources which had been designed from the past research. Radiation shielding modeling for treatment room used MCNPX software. The radiation shielding was concrete baryte on each side that includes coated borated polyethylene 2 cm thick and it is featured with a sliding door with dimensions 220 × 87 × 200 cm coated with stainless steels 2 cm thick. Results obtained value equivalent dose rate of neutron and gamma of each 41.5 µSv.h-1 and 2.05 µSv.h-1. Effects of radiation received by workers in the form of deterministic effects did not have a significant are impact

An Optimization Design of Collimator in The Thermal Column of Kartini Reactor For BNCT

Studies were carried out to design a collimator which results in epithermal neutron beam for in vivo experiment of Boron Neutron Capture Therapy (BNCT) at the Kartini Research Reactor by means of Monte Carlo N-Particle (MCNP) codes. Reactor within 100 kW of thermal power was used as the neutron source. All materials used were varied in size, according to the value of mean free path for each material. MCNP simulations indicated that by using 6 cm thick of Natural Nickel as collimator wall, 65 cm thick of Al as moderator, 3 cm thick of Ni-60 as filter, 6 cm thick of Bi as γ-ray shielding, 3.5 cm thick of Li2CO3-polyethilene, with 2 cm aperture diameter. Epithermal neutron beam with maximum flux of 6.60 x 108n.cm-2.s-1 could be produced. The beam has minimum fast neutron and γ-ray components of, respectively, 1.82 x 10-13Gy.cm2.n-1 and 1.70 x 10-13 Gy.cm2.n-1, minimum thermal neutron per epithermal neutron ratio of 0.041, and maximum directionality of 2,12. It did not fully pass the IAEA’s criteria, since the epithermal neutron flux was below the recommended value, 1.0 x 109 n.cm-2.s-1. Nonetheless, it was still usable with epithermal neutron flux exceeding 5.0 x 108 n.cm-2.s-1. it is still feasible for BNCT in vivo experiment

Internal Dose Analysis for Radiation Worker in Cancer Therapy Based on Boron Neutron Capture Therapy with Neutron Source Cyclotron 30 MeV Using Monte Carlo N Particle Extended Simulator

Based Studies were carried out to analyze internal dose for radiation worker at Boron Neutron Capture Therapy (BNCT) facility base on Cyclotron 30 MeV with BSA and room that actually design before. This internal dose analyze include interaction between neutron and air. The air contains N2 (72%), O2 (20%), Ar (0.93%), CO2, Neon, Kripton, Xenon, Helium and Methane. That internal dose to the worker should be bellow limit dose for radiation worker amount of 20 mSv/years. From the particle 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 flux Neutron in the air 3,16x107 Neutron/cm2s. room design in cancer facility have a measurement of length 200 cm, width 200 cm and high 166,40 cm. flux neutron can be used to calculated the reaction rate which is 80,1x10-2 reaction/cm3s for carbon-14 and 8,75x10-5 reaction/cm3s. Internal dose exposed to the radiation worker is 9.08E-9 µSv

Investigation on Neutronic Parameters of the KLT-40S Reactor Core with U3Si2-FeCrAl using SCALE Code

From a safety point of view, the fuel-cladding of the current design of the KLT-40S reactor still carries a potential risk in the event of a loss-of-coolant accident (LOCA) allowing the formation of hydrogen gas. The concept of accident tolerant fuels (ATF) offers a variety of new safer fuel-cladding materials, one of which is U3Si2-FeCrAl, a potential fuel-cladding combination according to various research sources. In this research, a study of neutronic parameters (1) cycle length, (2) reactivity feedback coefficient, and (3) reactor proliferation resistance was performed with ATF material U3Si2-FeCrAl as fuel-cladding in the KLT-40S reactor core. Modeling and simulation of the ATF-fueled KLT-40S reactor core were performed using KENO-VI and TRITON modules from SCALE code. The results showed that replacement of the fuel-cladding material with the ATF material in the KLT-40S reactor resulted in a shorter cycle length, and the enrichment required to reproduce the original cycle length was above the safeguard limit. The fuel temperature, moderator temperature, and void reactivity coefficient were negative, although not as negative as the original ones. The spent fuel produced at the end of the cycle had good proliferation resistance, although not as good as the original one

Transmutation of Transuranic Elements as Solid Coating in Molten Salt Reactor Fuel Channel

The accumulation of spent nuclear fuel (SNF) is presently considered as a hindrance of the massive deployment of nuclear power plant, especially regarding the transuranic (TRU) elements. Eliminating TRU through transmutation is one of the most feasible alternative as a technical solution to solve the issue. This study explores the possibility of TRU transmutation using molten salt reactor (MSR) in a heterogeneous configuration, where a solid TRU is coated inside the fuel channel filled with liquid salt fuel. Such configuration is proposed to allow higher TRU loading into fluoride salt mixture without compromising the safety of the reactor. TRU coating was applied in consecutively outward radial fuel channel layers with coating thicknesses of 2.5 mm and 5 mm. Calculation was performed using MCNP6.2 radiation transport code and ENDF/B-VII.0 neutron cross section library. From the results, TRU coating with smaller thickness and positioned closer to the centre of the core exhibit higher transmutation efficiency due to exposure to higher neutron flux. Highest transmutation efficiency was achieved at 67.93% after 160 days of burnup. This shows a potential of achieving highly efficient TRU using heterogeneous configuration in MSR core