• Nuclear Engineering and Technology
• /
• 제7권4호
• /
• pp.277-284
• /
• 1975

TRIGA Mark-III 원자로에 설치된 조사시설에서 중성자-감마 혼합 방사선장의 상대적 선량분포 특성을 중성자와 감다 방사선에 대한 감응함수가 다른 한쌍의 열형광선량계를 사용하여 측정한 결과이다. 수평방향 및 수직방향의 거리에 따른 비교적 균일한 선량 분포를 공유한 지역은 조사실 바닥으로부터 약 40cm와 130cm의 높이데 있는 두 수평 평행판사이의 구역증 조사실쪽으로 반원통형으로 돌출된 알미늄 저수조표면에서 수평방향으로 100cm이상의 거리에 있는 지역에 한정됨이 관찰되었다. 그 이외의 지역은 급격한 선량구배 특성을 갖고 있었고 특히 반원통형 알미늄수조 표면근처와 조사실 콩크리트 차폐내벽근처에서 더욱 구배가 컸다.

• 전기의세계
• /
• 제19권4호
• /
• pp.18-23
• /
• 1970

A neutron detector, in general, can not be utilized as the thermal neutron detecting chamber in the nuclear power reactor, especially P.W.R. due to the characteristics of high temperature, high pressure and high neutron flux in a reactor vessel. We have performed an experiment to detect the thermal neutrons at 400.deg. C and high flux of thermal neutron in a power reactor. Coating boron-10 on the aluminium plates by means of surface diffusion method at 600.deg. C for 5 hours in an electric furace, also we made a typical chamber which was compensated ionization chamber filled with free air as an ionization gas. It was checked the chamber characteristics in the TRIGA MARK-II Reactor at the power level from zero to 250KW. The chamber current showed a perfect linear increase to power increase. However, many variation of the measured current were observed within the power of 50KW.

• Nuclear Engineering and Technology
• /
• 제54권2호
• /
• 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.

• 전기의세계
• /
• 제15권5호
• /
• pp.1-5
• /
• 1966

The stability and safety of operation of a reactor is determined mainly by the sign and magnitude of its reactivity responses to temperature changes. Reactors are subject to temperature fluctuations due to the changes in reactor power and ambient temperature. These temperature fluctuations cause reactivity disturbances through changes in the nuclear and physical properties of the core. Because of these important phenomena by the temperature effects, a large portion of study and testing on a reactor design has been conducted. In this experiment the overall temperature coefficient of the TRIGA MARK-II reactor is measured. The basic procedure is to change the tgemperature of the water moderator, and from the movements of a newly recalibrated control rod(this is necessary due to the effects of fuel burn-up and control rod depression) required to mintain criticality, the reactivity worth of the temperature change is determined. From this measurement, the overall temperature coefficient seems to be smoothly varying, almost a linear function of temperature, and a value of approximately -0.267${\c}$/$^{\circ}C$ can be obtained for an average temperature range from $17.6^{\circ}C$ to $32.5^{\circ}C$.

• Nuclear Engineering and Technology
• /
• 제54권11호
• /
• pp.4226-4230
• /
• 2022

The Korea Atomic Energy Research Institute (KAERI) has designed a Hybrid-Low Power Research Reactor (H-LPRR) which can be used for critical assembly and conventional research reactor as well. It is an open tank-in-pool type research reactor (Thermal Power: 50 kW_th) of which the most important applications are Neutron Activation Analysis (NAA), Radioisotope (RI) production, education and training. There are eight irradiation holes on the edge of the reactor core: IR (6 holes for RI production) and NA (2 holes for NAA) holes. In order to quantify the elemental concentration in target samples through the Instrumental Neutron Activation Analysis (INAA), it is necessary to measure neutron spectrum parameters such as thermal neutron flux, the deviation from the ideal 1/E epithermal neutron flux distribution (α), and the thermal-to-epithermal neutron flux ratio (f) for the irradiation holes. In this study, the MCNP6.1 code and FORTRAN 90 language are applied to determine the parameters for the two irradiation holes (NA-SW and NA-NW) in H-LPRR, and in particular its α and f parameters are compared to values of other research reactors. The results confirmed that the neutron irradiation holes in H-LPRR are designed to be sufficiently applied to neutron activation analysis, and its performance is comparable to that of foreign research reactors including the TRIGA MARK II.

• Nuclear Engineering and Technology
• /
• 제54권2호
• /
• pp.608-616
• /
• 2022

Power peaking factors (PPF) is an important parameter for safe and efficient reactor operation. There are several methods to calculate the PPF at TRIGA research reactors such as MCNP and TRIGLAV codes. However, these methods are time-consuming and required high specifications of a computer system. To overcome these limitations, artificial intelligence was introduced for parameter prediction. Previous studies applied the neural network method to predict the PPF, but the publications using the ANFIS method are not well developed yet. In this paper, the prediction of PPF using the ANFIS was conducted. Two input variables, control rod position, and neutron flux were collected while the PPF was calculated using TRIGLAV code as the data output. These input-output datasets were used for ANFIS model generation, training, and testing. In this study, four ANFIS model with two types of input space partitioning methods shows good predictive performances with R² values in the range of 96%-97%, reveals the strong relationship between the predicted and actual PPF values. The RMSE calculated also near zero. From this statistical analysis, it is proven that the ANFIS could predict the PPF accurately and can be used as an alternative method to develop a real-time monitoring system at TRIGA research reactors.

• Journal of Radiation Protection and Research
• /
• 제32권4호
• /
• pp.190-193
• /
• 2007

The objective of this study is to recommend the radiation protection design parameters from the shielding point of view for concrete wall between the decay tank room and the primary pump room in TRIGA Mark-II Research Reactor Facility. The shield design for this concrete wall has been performed with the help of Point-kernel Shielding Code Micro-Shield 5.05 and this design was also validated based on the measured dose rate values with Radiation Survey Meter (G-M Counter) considering the ICRP-60 (1990) recommendations for occupational dose rate limit ($10{\mu}Sv/hr$). The recommended shield design parameters are: (i) thickness of 114.3 cm Ilmenite-Magnetite Concrete (IMC) or 129.54 cm Ordinary Reinforced Concrete (ORC) for concrete wall A (ii) thickness of 66.04 cm Ilmenite-Magnetite Concrete (IMC) or 78.74 cm Ordinary Reinforced Concrete (ORC) for concrete wall B and (iii) door thickness of 3.175 cm Mild Steel (MS) on the entrance of decay tank room. In shielding efficiency analysis, the use of I-M concrete in the design of this concrete wall shows that it reduced the dose rate by a factor of at least 3.52 times approximately compared to ordinary reinforced concrete.

• Nuclear Engineering and Technology
• /
• 제53권3호
• /
• pp.942-948
• /
• 2021

The aim of this work is to help inform the decision for choosing a convenient material for the PGAA (Prompt Gamma Activation Analysis) collimator plug to be installed at the tangential channel of the Moroccan Triga Mark II Research Reactor. Two families of materials are usually used for collimator construction: a mixture of high-density polyethylene (HDPE) with boron, which is commonly used to moderate and absorb neutrons, and heavy materials, either for gamma absorption or for fast neutron absorption. An investigation of two different collimator designs was performed using N-Particle Monte Carlo MCNP6.2 code with the ENDF/B-VII.1 and MCLIP84 libraries. For each design, carbon steel and lead materials were used separately as collimator heavy materials. The performed study focused on both the impact on neutron beam quality and the neutron-gamma background at the exit of the collimator beam tube. An analysis and assessment of the principal findings is presented in this paper, as well as recommendations.

• Nuclear Engineering and Technology
• /
• 제37권2호
• /
• pp.139-150
• /
• 2005

A robust control design procedure for a nuclear reactor has been developed and experimentally validated on the Penn State TRIGA research reactor. The utilization of the robust controller as a component of an autonomous control system is also demonstrated. Two methods of specifying a low order (fourth-order) nominal-plant model for a robust control design were evaluated: 1) by approximation based on the 'physics' of the process and 2) by an optimal Hankel approximation of a higher order plant model. The uncertainty between the nominal plant models and the higher order plant model is supplied as a specification to the ,u-synthesis robust control design procedure. Two methods of quantifying uncertainty were evaluated: 1) a combination of additive and multiplicative uncertainty and 2) multiplicative uncertainty alone. The conclusions are that the optimal Hankel approximation and a combination of additive and multiplicative uncertainty are the best approach to design robust control for this application. The results from nonlinear simulation testing and the physical experiments are consistent and thus help to confirm the correctness of the robust control design procedures and conclusions.

• Nuclear Engineering and Technology
• /
• 제48권1호
• /
• pp.200-210
• /
• 2016

The neutron radiography reactor (NRAD) is a 250 kW Mark-II Training, Research, Isotopes, General Atomics (TRIGA) reactor at Idaho National Laboratory, Idaho Falls, ID, USA. The East Radiography Station (ERS) is one of two neutron beams at the NRAD used for neutron radiography, which sits beneath a large hot cell and is primarily used for neutron radiography of highly radioactive objects. Additional fuel elements were added to the NRAD core in 2013 to increase the excess reactivity of the reactor, and may have changed some characteristics of the neutron beamline. This report discusses characterization of the neutron beamline following the addition of fuel to the NRAD. This work includes determination of the facility category according to the American Society for Testing and Materials (ASTM) standards, and also uses an array of gold foils to determine the neutron beam flux and evaluate the neutron beam profile. The NRAD ERS neutron beam is a Category I neutron radiography facility, the highest possible quality level according to the ASTM. Gold foil activation experiments show that the average neutron flux with length-to-diameter ratio (L/D) = 125 is $5.96{\times}10^6n/cm^2/s$ with a $2{\sigma}$ standard error of $2.90{\times}10^5n/cm^2/s$. The neutron beam profile can be considered flat for qualitative neutron radiographic evaluation purposes. However, the neutron beam profile should be taken into account for quantitative evaluation.