• 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＄.

• 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.

• Nuclear Engineering and Technology
• /
• v.11 no.1
• /
• pp.21-27
• /
• 1979

The relative reaction rates were measured in the TRIGA Mark II reactor core and analyzed to obtain the neutron spectrum parameters; relative neutron temperature T$^{n}$ and epithermal index (equation omitted) Measurements were made with the central thimble and the F2 position containing the light water. The relative neutron temperature was represented by the activation ratio of Lu-Mn, and the epithermal index was measured by Au-Mn foil activation. The multichannel analyzer was used to measure the relative ${\gamma}$-rays of the detector foils. The results were compared with the calculated values.

• 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.

• Nuclear Engineering and Technology
• /
• v.48 no.5
• /
• pp.1273-1279
• /
• 2016

Production of iodine-131 by neutron activation of tellurium in tellurium dioxide ($TeO_2$) material requires a target that meets the safety requirements. In a radiopharmaceutical production unit, a new lid for a can was designed, which permits tight sealing of the target by using tungsten inert gaswelding. The leakage rate of all prepared targets was assessed using a helium mass spectrometer. The accepted leakage rate is ${\leq}10^{-4}mbr.L/s$, according to the approved safety report related to iodine-131 production in the TRIGA Mark II research reactor (TRIGA: Training, Research, Isotopes, General Atomics). To confirm the resistance of the new design to the irradiation conditions in the TRIGA Mark II research reactor's central thimble, a study of heat effect on the sealed targets for 7 hours in an oven was conducted and the leakage rates were evaluated. The results show that the tightness of the targets is ensured up to $600^{\circ}C$ with the appearance of deformations on lids beyond $450^{\circ}C$. The study of heat transfer through the target was conducted by adopting a one-dimensional approximation, under consideration of the three transfer modes-convection, conduction, and radiation. The quantities of heat generated by gamma and neutron heating were calculated by a validated computational model for the neutronic simulation of the TRIGA Mark II research reactor using the Monte Carlo N-Particle transport code. Using the heat transfer equations according to the three modes of heat transfer, the thermal study of I-131 production by irradiation of the target in the central thimble showed that the temperatures of materials do not exceed the corresponding melting points. To validate this new design, several targets have been irradiated in the central thimble according to a preplanned irradiation program, going from4 hours of irradiation at a power level of 0.5MWup to 35 hours (7 h/d for 5 days a week) at 1.5MW. The results showthat the irradiated targets are tight because no iodine-131 was released in the atmosphere of the reactor building and in the reactor cooling water of the primary circuit.

• Nuclear Engineering and Technology
• /
• v.5 no.4
• /
• pp.334-338
• /
• 1973

The horizontal distribution of the fast neutron flux density in the Bulk Shielding Experimental Tank of the TRIGA Mark-II reactor at the steady power of 250 KW has been measured using a solid state track detector which is natural mica placed in contact with $^{232}$ Th fissile foil. The neutron flux density was calculated on the assumption that the fast neutron spectrum is similar to that from the thermal-induced $^{235}$ U fission. The resulting flux density distribution along the horizontal line from the center of the thermalizing column door is presented in tabular and graphical forms.

• Nuclear Engineering and Technology
• /
• v.55 no.10
• /
• pp.3732-3753
• /
• 2023

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the k_eff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.

• Nuclear Engineering and Technology
• /
• v.5 no.4
• /
• pp.291-309
• /
• 1973

Exposure rates due to leakage gamma-rays from operating reactors TRIGA Mark II and III were measured in a horizontal plane by means of scintillation spectrometry using a 3"$\times$3" cylindrical Nal(T1) detector associated with a 400 channel pulse height analyzer under varied conditions of reactor operation. In determining exposure rate due to the leakage gamma-rays at each point of measurement, Moriuchi's spectrum-exposure rate conversion theory was applied instead of using conventional responce matrix method which necessitates very complicated procedures to convert a spectrum into exposure rate. The results show that a basic pattern of "typical" spectrum of the reactor leakage gamma-rays is neither affected by thermal output of the reactor, nor influenced by overall attenuation in radiation intensity. It was indicated that he attenuation of the leakage gamma-rays in air in terms of exposure rate as a whole follows an exponential law, and the total exposure rate due to the leakage gamma-rays at a certain point is nearly proportional to thermal output of the reactor. The complexity in spectrum measured for a movable core reactor, TRIGA Mark III, was analyzed through spectrum resolution, and proper judgement of the leakage gamma-rays in a complex spectrum was discussed.ctrum was discussed.

• Nuclear Engineering and Technology
• /
• v.54 no.2
• /
• pp.523-531
• /
• 2022

In order to study the sensitivity and the uncertainty of the Moroccan research reactor TRIGA Mark II, a model of this reactor has been developed in our ERSN laboratory for use with the N-Particle MCNP Monte Carlo transport codes (version 6). In this article, the sensitivities of the effective multiplication factor of this reactor are evaluated using the ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 libraries and in 44 energy groups, for the cross sections of the fuel (U-235 and U-238) and the moderator (H-1 and O-16). However, the quantification of the uncertainty of the nuclear data is performed using the nuclear code NJOY99 for the generation and processing of covariance matrices. On the one hand, the highest uncertainty deviations, calculated using the ENDFB-VII.1 and JENDL4.0 evaluations, are 2275, 386 and 330 pcm respectively for the reactions U₂₃₅(n, f), $ U_{235}(n\bar{\nu})$ and H₁(n, γ). On the other hand, these differences are very small for the neutron reactions of O-16 and U-238. Regarding the neutron spectra, in CT-mid plane, they are very close for the three evaluations (ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0). These spectra present two peaks (thermal and fission) around the energies 0.05 eV and 1 MeV.