• Nuclear Engineering and Technology
• /
• v.51 no.2
• /
• pp.337-344
• /
• 2019

The TRIGA fuel of the Philippine Research Reactor-1 (PRR-1) will be used in a subcritical reactor assembly (SRA) to strengthen and advance nuclear science and engineering expertise in the Philippines. SRA offers a versatile and safe training and research facility since it can produce neutrons through nuclear fission reaction without achieving criticality. In this work, we used a geometrically detailed model of the PRR-1 TRIGA fuel to design a subcritical reactor assembly and calculate physical parameters of different fuel configurations. Based on extensive neutron transport simulations an SRA configuration is proposed, comprising 44 TRIGA fuel rods arranged in a $7{\times}7$ square lattice. This configuration is found to have a maximum $k_{eff}$ value of $0.95001{\pm}0.00009$ at 4 cm pitch. The SRA is characterized by calculating the 3-dimensional neutron flux distribution and neutron spectrum. The effective delayed neutron fraction and mean neutron generation time of the system are calculated to be $748pcm{\pm}7pcm$ and $41{\mu}s$, respectively. Results obtained from this work will be the basis of the core design for the subcritical reactor facility that will be established in the Philippines.

• Nuclear Engineering and Technology
• /
• v.10 no.2
• /
• pp.87-92
• /
• 1978

Multigroup constant calculation system for TRIGA-type reactor analysis was provided. Calculations for initial criticality, temperature coefficient, flux and power distributions of TRICA-Mark III reactor were carried out by using diffusion code CITATION. And some of results were compared with the values of start-up experiments and design values. It could be confirmed that the prepared computation system is very useful for TRIGA-type reactor analysis.

• Nuclear Engineering and Technology
• /
• v.50 no.1
• /
• pp.165-169
• /
• 2018

An assessment for determining N-16 activity concentrations during the operation condition of Bangladesh Atomic Energy Commission TRIGA Research Reactor was performed employing several governing equations. The radionuclide N-16 is a high energy (6.13 MeV) gamma emitter which is predominately created by the fast neutron interaction with O-16 present in the reactor core water. During reactor operation at different power level, the concentration of N-16 at the reactor bay region may increase causing radiation risk to the reactor operating personnel or the general public. Concerning the safety of the research reactor, the present study deals with the estimation of N-16 activity concentrations in the regions of reactor core, reactor tank, and reactor bay at different reactor power levels under natural convection cooling mode. The estimated N-16 activity concentration values with 500 kW reactor power at the reactor core region was $7.40{\times}10^5Bq/cm^3$ and at the bay region was $3.39{\times}10^5Bq/cm^3$. At 3 MW reactor power with active forced convection cooling mode, the N-16 activity concentration in the decay tank exit water was also determined, and the value was $4.14{\times}10^{-1}Bq/cm^3$.

• Nuclear Engineering and Technology
• /
• v.22 no.1
• /
• pp.58-65
• /
• 1990

A PC-based system for measuring and analysing random neutron process in the thermal reactor is developed and applied to TRIGA Mark-II reactor at KAERI. It is confirmed that this system has several advantages compared to conventional methods. So far, two techniques, autocorrelation and variance to mean ratio (VTMR), have been applied for analysing the count data collected from the single detector by using this system. The results of the two techniques agree within acceptable difference, but VTMR's results show much superior statistical reliability than those of autocorrelation especially when it is near critical. The $\beta$/Λ of TRIGA Mark-II reactor is measured to be about 125/sec when the reactivity is within -3＄ and about 150/sec when it is below -4＄.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2003.11a
• /
• pp.720-724
• /
• 2003

In order to develop the decontamination process for self-disposal with authorization of duct waste generated from the decommissioning of retired TRIGA research reactors, the surface characterization of duct specimen taken from TRIGA research reactor was carried out and the adequate decontamination method was selected. It can be known that the paint coated internal surface of duct is contaminated with $^{60}Co$and $^{137}Cs$, which are penetrated into the paint layer and incorporated into zinc plated surface of galvanized iron as the material of duct. Two step chemical decontamination process, in which sodium hydroxide and sulfuric acid solutions are used in turn, is quite successful to remove the surface contamination of duct waste.

• Nuclear Engineering and Technology
• /
• v.50 no.1
• /
• pp.35-42
• /
• 2018

This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics) Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4) and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE) codes. Cores 133 and 134 were analyzed in 2-D (r, ${\theta}$) and 3-D (r, ${\theta}$, z), using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0), Joint Evaluated Fission and Fusion File (JEFF-3.1), Japanese Evaluated Nuclear Data Library (JENDL-3.2), and Joint Evaluated File (JEF-2.2) nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.10a
• /
• pp.276.1-276
• /
• 1999

• Nuclear Engineering and Technology
• /
• v.11 no.4
• /
• pp.287-293
• /
• 1979

Mixed standard-FLIP fuel loading patterns in the TRIGA Mark-III reactor were analyzed. It was judged that the mixed loading pattern with the standard fuel in the B-ring and the FLIP fuel in other rings was mostly desirable in view of fuel temperature, cooling condition with the natural convection, or effective thermal flux utilization in the central thimble. In addition, tile maximum useful flux in tile reactor beamports versus the loading patterns was evaluated.

• Nuclear Engineering and Technology
• /
• v.2 no.2
• /
• pp.67-72
• /
• 1970

The measurements of the fast neutron flux and its spectrum have been carried out by the threshold detectors in the central thimble of TRIGA Mark-II reactor operating at 250 KW. The following reactions have been employed for these measurements, viz : Ni$^{58}$ (n, p) Co$^{58}$ ； $Mg^{24}$ (n, p) Na$^{24}$ ； $Al^{27}$ (n, $\alpha$) Na$^{24}$ . From the activation data the fast neutron spectrum were calculated by CDC-3600 computer making use of two semi-empirical methods. It has been verified that the validity of assumption of a fission spectrum in the central thimble exists only above 1 to 2 Mev energy level. With this spectrum, a fast neutron flux in the range of 1 $\times$ 10$^{12}$ n/$\textrm{cm}^2$-sec above the energy of 2.6 Mev was observed in the central thimble of TRIGA MARK-II reactor.

• Nuclear Engineering and Technology
• /
• v.3 no.4
• /
• pp.185-197
• /
• 1971

Korea's TRIGA Mark-Ⅱ reactor was primarily designed in 1950's and was constructed in 1962 for 100 kw thermal output, but it was upgraded to 250 kw in July 1969. Nevertheless, the shield remains unchanged, although the radiation level has increased. The result of computation On this paper shows that, with the existing shield, it is safe for the fast neutrons even after the power upgrading by 2.5 times. It is, however, somewhat dangerous for the gamma rays which are comprised of primary and secondary. For the analysis of the reactor shielding design, an attempt is made for the computation toward the horizontal direction. From theoretical point of view, it can be concluded that some layer of additional shield must be reinforced to the existing concrete in order to be radiologically safe in the reactor hall.