scholarly journals DOSE ESTIMATION OF THE BNCT WATER PHANTOM BASED ON MCNPX COMPUTER CODE SIMULATION

2020 ◽  
Vol 22 (1) ◽  
pp. 23
Author(s):  
Amanda Dhyan Purna Ramadhani ◽  
Susilo Susilo ◽  
Irfan Nurfatthan ◽  
Yohannes Sardjono ◽  
Widarto Widarto ◽  
...  
Keyword(s):  
Cancer Cell ◽  
Human Body ◽  
Research Reactor ◽  
Irradiation Time ◽  
Computer Code ◽  
Dose Estimation ◽  
Water Phantom ◽  
Cubic Form ◽  
Epithermal Neutrons ◽  
Cylindrical Form

Cancer is a malignant tumor that destroys healthy cells. Cancer treatment can be done by several methods, one of which is BNCT. BNCT uses 10B target which is injected into the human body, then it is irradiated with thermal or epithermal neutrons. Nuclear reaction will occur between boron and neutrons, producing alpha particle and lithium-7. The dose is estimated by how much boron and neutron should be given to the patient as a sum of number of boron, number of neutrons, number of protons, and number of gamma in the reaction of the boron and neutron. To calculate the dose, the authors simulated the reaction with Monte Carlo N Particle-X computer code. A water phantom was used to represent the human torso, as 75% of human body consists of water. Geometry designed in MCNPX is in cubic form containing water and a cancer cell with a radius of 2 cm. Neutron irradiation is simulated as originated from Kartini research reactor, modeled in cylindrical form to represent its aperture. The resulting total dose rate needed to destroy the cancer cell in GTV is 2.0814×1014 Gy.s (76,38%) with an irradiation time of 1,4414×10-13 s. In PTV the dose is 5.2295×1013 Gy.s (19,19%) with irradiation time of 5.7367×10-13 s. In CTV, required dose is 1.1866×1013 Gy.s (4,35%) with an irradiation time of 2.5283×10-12 s. In the water it is 1.9128×1011 Gy.s (0,07%) with an irradiation time of 1,5684×10-10 s. The irradiation time is extremely short since the modeling is based on water phantom instead of human body.Keywords: BNCT, Dose, Cancer, Water Phantom, MCNPX

scholarly journals Fuel burnup analysis for Thai research reactor by using MCNPX computer code

2017 ◽  
Vol 860 ◽  
pp. 012033 ◽  
Author(s):  
S Sangkaew ◽  
T Angwongtrakool ◽  
B Srimok
Keyword(s):  
Research Reactor ◽  
Computer Code ◽  
Fuel Burnup

scholarly journals The radioactivity analysis of14C and3H in graphite from the dismantled korea research reactor and its dose estimation

2009 ◽  
Vol 44 (5) ◽  
pp. 59-63
Author(s):  
H. R. Kim ◽  
S. B. Hong ◽  
W. Lee ◽  
G. H. Chung ◽  
G. S. Choi ◽  
...  
Keyword(s):  
Research Reactor ◽  
Dose Estimation ◽  
Radioactivity Analysis

scholarly journals OPTIMIZATION OF COLLIMATOR APERTURE GEOMETRY FOR BNCT KARTINI RESEARCH REACTOR USING MCNPX

2019 ◽  
Vol 21 (1) ◽  
pp. 9
Author(s):  
Ramadhan Valiant Gill S.B. ◽  
Yohannes Sardjono
Keyword(s):  
Neutron Source ◽  
Research Reactor ◽  
Irradiation Time ◽  
Base Point ◽  
Cell Diameter ◽  
Cylindrical Shape ◽  
Trace Distance ◽  
Optimal Output ◽  
Boron Neutron Capture ◽  
High Ionization

Boron Neutron Capture Therapy (BNCT) is one of the promising cancer therapy modalities due to its selectivity which only kills the cancer cells and does not damage healthy cells around cancer. In principle, BNCT utilizes the high ionization properties of alpha (4He) and lithium (7Li) particles derived from the reaction between epithermal and boron-10 neutrons (10B + n → 7Li + 4He) in cells, where trace distance of alpha and lithium particles is equivalent with cell diameter. The neutron source used in BNCT can come from a reactor, as a condition for conducting BNCT therapy tests, there are five standard parameters that must be met for a neutron source to be used as a source, and the standards come from IAEA. This research is based on simulation using the MCNPX program which aims to optimize IAEA parameters that have been obtained in previous studies by changing the shape of the collimator geometry from cone shape to cylinder with variations diameter from 3, 5 and 10 cm and also the simulation divided into two schemes namely first moderator Al is placed in a position 9.5 cm behind the collimator and the second is the moderator Al is pressed into the base point of the aperture in the collimator. In this work, neutrons originated from Yogyakarta Kartini research reactor have the energy range in the continuous form. The results of the optimization on each scheme of the collimator are compared with the outputs that have been obtained in previous studies where the aperture of the collimator is in the cone shape. The most optimal output obtained from the results is a collimator with a diameter of 5 cm in the second scheme where the results of IAEA parameters that are produced (n/cm2 s) = 2.18E+8, / (Gy-cm2/n) = 6.69E-13, / (Gy-cm2/n) = 2.44E-13,  = 4.03E-01, and J/ = 6.31E-01. These results can still be used for BNCT experiments but need a long irradiation time and when compared to previous studies, the output of the collimator with the diameter of 5 cm is more optimal.Keywords: BNCT, Collimator, IAEA Parameters, MCNPX, Cylindrical shape 

scholarly journals Evaluation of technological radiation sources in the pumping station of the primary circulation loop and on the top of the pool type research reactor

2018 ◽  
Vol 33 (1) ◽  
pp. 31-46
Author(s):  
Stoyan Kadalev
Keyword(s):  
Heat Carrier ◽  
Decay Time ◽  
Research Reactor ◽  
Irradiation Time ◽  
Nuclear Research ◽  
Circulation Loop ◽  
Nuclear Facilities ◽  
Energy Agency ◽  
Radiation Sources ◽  
Pool Type

The present paper considers the approach to an assessment of technological radiation sources in the primary water-water reactor circulation loop. In principle, such an evaluation is a multidisciplinary task that covers not only the irradiation of the nuclei, the formation of new isotopes and their decay when they are unstable, but also calculations in the field of hydraulics in order to perform an assessment of the irradiation time and the decay time. A general and a more detailed review of the radiation sources formation in the nuclear facilities and the pool type research reactors with demineralized water as a heat carrier are prepared. The initial isotopic composition of the heat carrier has been adopted according to the Vienna Standard Mean Ocean Water recommended by the International Atomic Energy Agency. The general mathematical model of the processes of nuclei irradiation, the formation of new isotopes and their decay, the assessment of the irradiation time and the decay time is described in details, enabling the repetition of this evaluation to a particular facility. The presented approach is applied in the reconstruction design of the nuclear research reactor IRT-2000, Sofia, Bulgaria.

Neutron fluence depth profiles in water phantom on epithermal beam of LVR-15 research reactor

2010 ◽  
Vol 68 (4-5) ◽  
pp. 617-619
Author(s):  
L. Viererbl ◽  
V. Klupak ◽  
Z. Lahodova ◽  
M. Marek ◽  
J. Burian
Keyword(s):  
Research Reactor ◽  
Neutron Fluence ◽  
Depth Profiles ◽  
Water Phantom

scholarly journals OPTIMIZATION OF FUEL RELOADING PATTERNS FOR A RESEARCH REACTOR BY SIMULATED ANNEALING

2011 ◽  
Vol 14 (1) ◽  
pp. 63-71
Author(s):  
Binh Quang Do
Keyword(s):  
Simulated Annealing ◽  
Optimization Problem ◽  
Research Reactor ◽  
Diffusion Theory ◽  
Computer Code ◽  
Decision Variable ◽  
Safety Parameters ◽  
Fortran Language ◽  
Dalat Research Reactor ◽  
Effective Multiplication

This article presents results obtained from a research into an application of simulated annealing method to the in-core fuel reloading pattern optimization for a research reactor. The decision variable of the optimization problem is a fuel reloading pattern for the next cycle after the present cycle finishes. The objective function maximizes the effective multiplication factor keff at the beginning of cycle while it is established to include an important safety paramater – the power peaking factor, in search process. A procedure for searching the optimal solutions was formed and a computer code was developed in the Fortran language running on PCs. Nuclear safety parameters for the optimization problem are provided from the results of the multigroup neutron diffusion theory computation program CITATION. A sample calculation was performed to find the optimal fuel reloading patterns for the second cycle of the Dalat research reactor and the results are presented in this article.

scholarly journals Initial Criticality Analysis of Malaysia’s TRIGA Research Reactor Using TRIGLAV Computer Code

2018 ◽  
Vol 7 (4.35) ◽  
pp. 899
Author(s):  
M. A Khattak ◽  
Abdoulhdi A. Borhana ◽  
Nurul Syahrizzat M. Yasin ◽  
Rustam Khan ◽  
Juniza Md Saad
Keyword(s):  
Diffusion Equation ◽  
Power Distribution ◽  
Research Reactor ◽  
Nuclear Research ◽  
Computer Code ◽  
Experimental Result ◽  
Zirconium Hydride ◽  
Finite Differences Method ◽  
Triga Mark Ii ◽  
Criticality Analysis

Malaysian Nuclear Agency hosts the 1MW Thermal TRIGA MARK II research reactor since 1982. It first initial criticality was achieved on 28th June 1982 with loading of solid fuel elements like Uranium Zirconium Hydride. TRIGA MARK II started its operation of the research reactor on the same year with a 1MW power generation. Training, Research, Isotope Production and General Atomic (TRIGA) is designed to successfully actualize the variety fields of fundamental nuclear research, the manpower training and the production of radioisotopes. This study deals with the initial criticality analysis of the TRIGA research reactor using TRIGLAV reactor physics computer program. For this purpose, a model of its initial core will be developed and simulated using the software and the results will be validated against the experimental result as mentioned in the final safety analysis report (FSAR). The TRIGLAV computer code solves the neutron diffusion equation by using a finite differences method with iteration of fission density. TRIGLAV is based on four group time independent diffusion equation in two dimensional cylindrical (r, θ) geometry.  TRIGLAV can also be applied for criticality of reactor, fuel burn-up calculations, and power distribution and flux distributions calculations of the core and also the reactivity predictions of the reactor.  

scholarly journals Radiological Assessment of Atmospheric Release from Tehran Research Reactor during Normal Operation by Using Pc-Cream Code

2016 ◽  
Vol 54 ◽  
pp. 17-26
Author(s):  
Nahid Sadeghi ◽  
Rohollah Ahangari
Keyword(s):  
Research Reactor ◽  
Computer Code ◽  
Laboratory Analysis ◽  
Air Transport ◽  
Normal Operation ◽  
Radiological Assessment ◽  
Tehran Research Reactor ◽  
Tehran Province ◽  
The Individual ◽  
Atmospheric Release

In this work, radiological assessment of atmospheric release from Tehran’s Research Reactor (TRR) stack and assessment of public exposures under normal operation has been studied. To perform tasks mentioned above, Pc-Cream computer code which simulates Gaussian Dispersion air transport plume model as well as laboratory analysis of the soil and leaves samples and TLD (Thermo Luminescent Dosimeter) monitoring around the TRR site was used. Results of the Pc-Cream code showed that the annual committed and external dose received by the individual in the vicinity of the reactor is below the regulatory limit. Also, the results of laboratory analysis of available radionuclides in the soil and leaves samples showed that the concentrations are close to the background (K40=635, Th232=28, Cs137=0.29 up to 28.82, Ra226=25 (Bq[1]/Kg) in soil and K40=457, Be7≈70 (Bq/Kg) in leaves) and confirm the code results. The monitored dose values of the TLD detectors were positioned around the reactor within 500 m radius shows that the background dose in vicinity of TRR (113 μSv up to 150 μSv) is consistent with the background dose in Tehran province (125 μSv).

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