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

Pellet-Cladding Interaction of LMFBR Fuel Elements at Unsteady State: An Introduction to Computer Code ISUNE-5

1981 ◽  
Vol 103 (4) ◽  
pp. 627-636 ◽  
Author(s):  
B. M. Ma
Keyword(s):  
Reactor Core ◽  
Fatigue Cracking ◽  
Computer Code ◽  
Fuel Burnup ◽  
Fuel Pellet ◽  
Gas Release ◽  
Unsteady State ◽  
Fission Gas ◽  
Fuel Elements ◽  
Fission Gas Release

The fuel pellet-cladding interaction (PCI) of liquid-metal fast breeder reactor (LMFBR) fuel elements or fuel rods at unsteady state is analyzed and discussed based on experimental results. In the analyses, the heat generation, fuel restructuring, temperature distribution, gap conductance, irradiation swelling, irradiation creep, fuel burnup, fission gas release, fuel pellet cracking, crack healing, cladding cracking, yield failure and fracture failure of the fuel elements are taken into consideration. To improve the sintered (U,Pu)O2 fuel performance and reactor core safety at high temperature and fuel burnup, it is desirable to (a) increase and maintain the ductility of cladding material, (b) provide sufficient gap thickness and plenum space for accommodating fission gas release, (c) keep ramps-power increase rate slow and gentle, and (d) reduce the intensity and frequency of transient PCI in order to avoid intense stress fatigue cracking (SFC) and stress corrosion cracking (SCC) due to fission product compounds CsI, CdI2, Cs2Te, etc. at the inner cladding surface of the fuel elements during PCI.

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 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 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).

Long.Term Reactivity Change and Control: On.Power Refuelling

2021 ◽  
pp. 33-44
Author(s):  
Wei Shen ◽  
Benjamin Rouben
Keyword(s):  
Residence Time ◽  
Thermal Flux ◽  
Computer Code ◽  
Fuel Burnup ◽  
The Core ◽  
Time Integral ◽  
Fuel Bundle ◽  
Computer Codes ◽  
Neutron Exposure ◽  
And Control

There are 2 concepts related to the “age” of fuel: irradiation (fluence) and fuel burnup. The fuel irradiation in a given fuel bundle, denoted ω, is defined as the time integral of the thermal flux in the fuel during its residence time in the core. Another term for irradiation is fluence. Irradiation is also known as the thermal-neutron exposure of the fuel. The units of irradiation are neutrons/cm2, or more conveniently, neutrons per kilobarn, n/kb. Since the cut-off of the thermal-energy range may be defined differently in different computer codes, the fuel irradiation may vary from computer code to computer code, and caution must therefore be exercised when comparing irradiation values using different codes. In documents, it has been more and more usual to report values of fuel burnup rather than fuel irradiation, as burnup does not suffer from differences in definition between codes.

Calculation of fuel burnup and excess reactivity using TRIGLAV code for the BAEC TRIGA research reactor

2020 ◽  
Vol 14 (4) ◽  
pp. 291
Author(s):  
Md. Mizanur Rahman ◽  
Anisur Rahman ◽  
Shafiul Hossain ◽  
P.K. Das ◽  
Md. Ashraf Ali ◽  
...  
Keyword(s):  
Research Reactor ◽  
Fuel Burnup

Validation of the Onsite Used Operational Code Against Burnup Measurement

2016 ◽  
Vol 3 (1) ◽  
Author(s):  
Michal Koleška ◽  
Michal Šunka ◽  
Jaroslav Ernest
Keyword(s):  
Experimental Data ◽  
Fission Product ◽  
Research Reactor ◽  
Fuel Burnup ◽  
Spent Fuel ◽  
Axial Distribution ◽  
Operational Code ◽  
Spectrometric System ◽  
Type Fuel

A spectrometric system was developed for spent fuel burnup evaluations at the LVR-15 research reactor, which employed highly enriched (36%) IRT-2M-type fuel. Such a system allows the measurement of fission product axial distribution by measuring certain nuclides, such as Cs137, Cs134, and their ratios, respectively. Within the paper, a comparison between experimental data provided by the spectrometric system and calculations in operational code called NODER is provided.

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