• Proceedings of the KIEE Conference
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• 2003.11c
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• pp.789-792
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• 2003

pMOSFET having 10 ${\mu}um$ thickness Gd layer has been tested to be used as a slow neutron sensor. The total thermal neutron cross section for the Gd is 47,000 barns and the cross section value drops rapidly with increasing neutron energy. When slow neutrons are incident to the Gd layer, the conversion electrons are emitted by the neutron absorption process. The conversion electrons generate electron-hole pairs in the $SiO_2$ layer of the pMOSFET. The holes are easily trapped in Oxide and act as positive charge centers in the $SiO_2$ layer. Due to the induced positive charges, the threshold turn-on voltage of the pMOSFET is changed. We have found that the voltage change is proportional to the accumulated slow neutron dose, therefore the pMOSFET having a Gd nuclear reaction layer can be used for a slow neutron dosimeter. The Gd-pMOSFET were tested at HANARO neutron beam port and $^{60}CO$ irradiation facility to investigate slow neutron response and gamma response respectively. Also the pMOSFET without Gd layer were tested at same conditions to compare the characteristics to the Gd-pMOSFET. From the result, we have concluded that the Gd-pMOSFET is very sensitive to the slow neutron and can be used as a slow neutron dosimeter. It can also be used in a mixed radiation field by subtracting the voltage change value of a pMOSFET without Gd from the value of the Gd-pMOSFET.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.427-432
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• 1997

Boron neutron capture therapy(BNCT) is a binary treatment modality that can selectively irradiate tumor tissue. More is known now about the radiation biology of BNCT, which has reemerged as a potentially useful method for preferential irradiation of tumors. We design a square reactor (that can easily be reconfigured into polygonal reactors as the need arises) with four slab type assemblies to produce high quality epithermal neutron beans and thermal neutron beams jot use in neutron capture therapy. With a low operating power of 300kW, the heat generated in the core can be removed by natural convection through a pool of tight water. The proposed design in this study could be constructed for a dedicated clinical BNCT facility that would operate very safely.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.787-791
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• 2017

Americium-beryllium (Am-Be; n, ${\gamma}$) is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci), yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources) experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

• Nuclear Engineering and Technology
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• v.7 no.4
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• pp.277-284
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• 1975

A result of measurement for the relative dose distribution of neutron gamma mixed radiation field in the biological irradiation facility installed at TRIGA Mark-III reactor is described. The relative dose distributions of neutron-gamma mixed radiation field in the biological exposure room have been experimentally determined using a thermoluminescent dosimeter. Presented herein in graphical forms are the experimental results obtained. It as observed that the region commonly having the characteristics of rather homogeneous horizontal and lateral dose distributions is confined to the area bounded by the two planes horizontally parallel to the beam direction with heights of about 40 cm and 130 cm, respectively, at distances beyond 100 cm from the segmentary surface of the aluminum pool liner projected into the the exposure room, while other areas show a steeper gradient in dosage, especially the places adjacent to the segment of the aluminum pool liner and near the inner po${\gamma}$lion of the concrete walls of the exposure room.

• Journal of the Korean Society of Mechanical Technology
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• v.13 no.2
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• pp.137-143
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• 2011

Fuel Test Loop(FTL) is a facility which could conduct a fuel irradiation test at HANARO(High-flux Advanced Neutron Application Reactor). FTL simulates commercial NPP's operating conditions such as the pressure, temperature and neutron flux levels to conduct the irradiation and thermo-hydraulic tests. The In-Pile Test Section(IPS) installed in HANARO FTL is designed as a pressure vessel design conditions of $350^{\circ}C$, 17.5MPa. The instrumentation MI-cables for thermocouples, SPND and LVDT are passed through the sealing plug, which is in the pressure boundary region and is a part of instrumentation feedthrough of MI-cable. In this study, the brazing method and performance test results are introduced to the sealing plug with BNi-2 filler metal, which is selected with consideration of the compatibility for the coolant. The performance was verified through the insulation resistance test, hydrostatic test, and helium leak test.

• Journal of Radiation Protection and Research
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• v.42 no.3
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• pp.141-145
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• 2017

Background: The concrete walls inside the vaults of cyclotron facilities are activated by neutrons emitted by the targets during radioisotope production. Reducing the amount of radioactive waste created in such facilities is very important in case they are decommissioned. Thus, we proposed a strategy of reducing the neutron activation of the concrete walls in cyclotrons during operation. Materials and Methods: A polyethylene plate and B-doped Al sheet (30 wt% of B and 2.5 mm in thickness) were placed in front of the wall in the cyclotron room of a radioisotope production facility for pharmaceutical use. The target was Xe gas, and a Cu block was utilized for proton dumping. The irradiation time, proton energy, and beam current were 8 hours, 30 MeV, and $125{\mu}A$, respectively. To determine a suitable thickness for the polyethylene plate set in front of the B-doped Al sheet, the neutron-reducing effects achieved by inserting such sheets at several depths within polyethylene plate stacks were evaluated. The neutron fluence was monitored using an activation detector and 20-g on de Au foil samples with and without 0.5-mm-thick Cd foil. Each Au foil sample was pasted onto the center of a polyethylene plate and B-doped Al sheet, and the absolute activity of one Au foil sample was measured as a standard using a Ge detector. The resulting relative activities were obtained by calculating the ratio of the photostimulated luminescence of each foil sample to that of the standard Au foil. Results and Discussion: When the combination of a 4-cm-thick polyethylene plate and B-doped Al sheet was employed, the thermal neutron rate was reduced by 78%. Conclusion: The combination of a 4-cm-thick polyethylene plate and B-doped Al sheet effectively reduced the neutron activation of the investigated concrete wall.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.941-950
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• 2013

The Korean Nuclear-Hydrogen Technology Development (NHTD) Plan will be performing irradiation testing of coated particle fuel at HANARO to support the development of VHTR in Korea. This testing will be carried out to demonstrate and qualify TRISO-coated particle fuel for use in VHTR. The testing will be irradiated in an inert gas atmosphere without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one has nine fuel compacts and the other five compacts and eight graphite specimens. Each compact contains about 260 TRISO-coated particles. The irradiation device is being loaded and irradiated into the OR5 hole of the in HANARO core from August 2013. The device will be operated for about 150 effective full-power days at a peak temperature of about $1030^{\circ}C$ in BOC (Beginning of Cycle) during irradiation testing. After a peak burn-up of about 4 atomic percentage and a peak fast neutron fluence of about $1.7{\times}10^{21}\;n/cm^2$, PIE (Post-Irradiation Examination) of the irradiated coated particle fuel will be performed at IMEF (Irradiated Material Examination Facility). This paper reviews the design of test rod and irradiation device for coated particle fuel, and discusses the technical results for irradiation testing at HANARO.

• Progress in Medical Physics
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• v.18 no.2
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• pp.87-92
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• 2007

A thermal neutron beam facility utilizing a typical tangential beam port for Neutron Capture Therapy was installed at the HANARO, 30 MW multi-purpose research reactor. Mixed beams with different physical characteristics and relative biological effectiveness would be emitted from the BNCT irradiation facility, so a quantitative analysis of each component of the mixed beams should be performed to determine the accurate delivered dose. Thus, various techniques were applied including the use of activation foils, TLDs and ionization chambers. All the dose measurements were perform ed with the water phantom filled with distilled water. The results of the measurement were compared with MCNP4B calculation. The thermal neutron fluxes were $1.02E9n/cm^2{\cdot}s\;and\;6.07E8n/cm^2{\cdot}s$ at 10 and 20 mm depth respectively, and the fast neutron dose rate was insignificant as 0.11 Gy/hr at 10 mm depth in water The gamma-ray dose rate was 5.10 Gy/hr at 20 mm depth in water Good agreement within 5%, has been obtained between the measured dose and the calculated dose using MCNP for neutron and gamma component and discrepancy with 14% for fast neutron flux Considering the difficulty of neutron detection, the current study support the reliability of these results and confirmed the suitability of the thermal neutron beam as a dosimetric data for BNCT clinical trials.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.469-476
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• 2020

Research reactors in-core experimental facilities are designed to provide the highest steady state flux for user's irradiation requirements. However, fuel conversion from highly enriched uranium (HEU) to low enriched uranium (LEU) driven by the ongoing effort to diminish proliferation risk, will impact reactor physics parameters. Preserving the reactor capability to produce the needed flux to perform its intended research functions, determines the conversion feasibility. This study investigates the neutron flux in the central experimental facility of two material test reactors (MTR), the IAEA generic10 MW benchmark reactor and the 22 MW s Egyptian Test and Research Reactor (ETRR-2). A 3D full core model with three uranium enrichment of 93%, 45%, and 20% was constructed utilizing the OpenMC particle transport Monte Carlo code. Neutronics calculations were performed for fresh fuel, the beginning of life cycle (BOL) and end of life cycle (EOL) for each of the three enrichments for both the IAEA 10 MW generic reactor and core 1/98 of the ETRR-2 reactor. Criticality calculations of the effective multiplication factor (K_eff) were executed for each of the twelve cases; results show a reasonable agreement with published benchmark values for both reactors. The thermal, epithermal and fast neutron fluxes were tallied across the core, utilizing the mesh tally capability of the code and are presented here. The axial flux in the central experimental facility was tallied at 1 cm intervals, for each of the cases; results for IAEA 10 MW show a maximum reduction of 14.32% in the thermal flux of LEU to that of the HEU, at EOL. The reduction of the thermal flux for fresh fuel was between 5.81% and 9.62%, with an average drop of 8.1%. At the BOL the thermal flux showed a larger reduction range of 6.92%-13.58% with an average drop of 10.73%. Furthermore, the fission reaction rate was calculated, results showed an increase in the peak fission rate of the LEU case compared to the HEU case. Results for the ETRR-2 reactor show an average increase of 62.31% in the thermal flux of LEU to that of the HEU due to the effect of spectrum hardening. The fission rate density increased with enrichment, resulting in 34% maximum increase in the HEU case compared to the LEU case at the assemblies surrounding the flux trap.

• Journal of Radiation Protection and Research
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• v.48 no.1
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• pp.52-57
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• 2023

Background: Korea Atomic Energy Research Institute (KAERI) operates several nuclear research facilities licensed by Nuclear Safety and Security Commission (NSSC). The emergency preparedness requirements, GSR Part 7, by International Atomic Energy Agency (IAEA) request protection strategy based on the hazard assessment that is not applied in Korea. Materials and Methods: In developing the protection strategy, it is important to consider an accident scenario and its consequence. KAERI has tried the hazard assessment based on a hypothesis accident scenario for the major nuclear facilities. During the assessment, the safety analysis report of the related facilities was reviewed, the simulation using MELCOR, MACCS2 code was implemented based on a considered accident scenario of each facility, and the international guidance was considered. Results and Discussion: The results of the optimized protective actions were 300 m evacuation and 800 m sheltering for the High-Flux Advanced Neutron Application Reactor (HANARO), the evacuation to radius 50 m, the sheltering 400 m for post-irradiation examination facility (PIEF), 100 m evacuation or sheltering for HANARO fuel fabrication plant (HFFP) facility. Conclusion: The results of the optimized protective actions and its distances for the KAERI facilities for the maximum postulated accidents were considered in establishing the emergency plan and procedures and implementing an emergency exercise for the KAERI facilities.