• Nuclear Engineering and Technology
• /
• v.51 no.7
• /
• pp.1715-1720
• /
• 2019

Jozef Stefan Institute (JSI), TRIGA Mark II reactor employs the homogeneous mixture of uranium and zirconium hydride fuel type. Since its upgrade, a series of fresh fuel steady state experimental benchmarks have been conducted. The benchmark results have provided data for testing computational neutronics codes which are important for reactor design and safety analysis. In this work, we investigated the JSI TRIGA Mark II reactor neutronics characteristics: the effective multiplication factor and two safety parameters, namely the control rod worth and the fuel temperature reactivity coefficient using SuperMC. The modeling and real-time cross section generation methods of SuperMC were evaluated in the investigation. The calculation analysis indicated the following: the effective multiplication factor was influenced by the different cross section data libraries; the control rod worth evaluation was better with Monte Carlo codes; the experimental fuel temperature reactivity coefficient was smaller than calculated results due to change in water temperature. All the results were in good agreement with the experimental values. Hence, SuperMC could be used for the designing and benchmarking of other TRIGA Mark II reactors.

• 전기의세계
• /
• v.14 no.6
• /
• pp.8-13
• /
• 1965

The TRIGA Mark-II Reactor is very simple to analyze the dynamic characteristics, so that the heat transfer function of the reactor fuel rod is able to be considered as a over-all feedback transfer function. The heat transfer dynamics of the fuel rod is derived under some assumptions. And the over-all reactor transfer function is analytically calcu- lated and it is compared with the measured value. The reactor dynamics and the stability are analyzed by means of the Root-Locus and the Nyquist.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.05a
• /
• pp.414-414
• /
• 1999

• 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.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.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2004.06a
• /
• pp.117-118
• /
• 2004

국내 TRIGA Mark II&III 및 우라늄 변환시설 해체시 다량의 금속폐기물 발생이 예상되며, 현재 이러한 저준위 방사성 금속폐기물을 처리, 재활용하는 기술들에 대한 연구가 국$\cdot$내외적으로 활발히 수행 중에 있다. 이에 본 연구에서는 해체시 발생하는 금속폐기물의 효율적인 감용 및 재활용에 관한 국내외 처리 기술을 검토하여 국내 TRIGA Mark II&III 및 우라늄 변환시설에서 발생하는 해체 금속폐기물의 처리 방안을 수립하기 위한 정보를 제공하고자 한다.

• Nuclear Engineering and Technology
• /
• v.2 no.1
• /
• pp.45-61
• /
• 1970

• 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.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.

• Journal of Radiation Protection and Research
• /
• v.24 no.2
• /
• pp.109-120
• /
• 1999

Residual radiation and radioactivity level in TRIGA Mark-II, III research reactors and facilities at the KAERI Seoul site, which are to be decommissioned, have been measured, analyzed and evaluated to know the current status of radiation and radioactivity level and to establish and to provide the technical requirements for the safe decommissioning of the facilities which shall be applied in minimizing the radiation exposure for workers and in preventing the release of the radioactive materials to the environment. Radiation dose rate and surface radioactivity contamination level on the experimental equipments, floors, walls of the facilities, and the surface of the activated materials within the reactor pool structure were measured and evaluated. Radioactivity and radionuclides in the pool and cooling water were also analyzed. In case of the activated reactor pool structures which are very difficult to measure the radiation and radioactivity level, a computer code Fispin was additionally used for estimation of the residual radioactivity and radionuclides. The radiation and radioactivity data obtained in this study were effectively used as basic data for decontamination and dismantling plan for safe decommissioning of TRIGA Mark-II, III facilities.