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The neutron flux distribution in the core region of the JRTR was measured through the neutron activation analysis method. Sample mount and a tool for withdrawing the sample mount were devised to irradiate the Au wires and foils. The measured data were compared with calculated values to confirm the reliability of the power distribution calculation. The reactor power during irradiation was estimated to be 9% lower than indications of power monitoring systems calibrated at around 10 μW. The measured transversal and axial neutron reaction rate distributions showed agreement within 1.32% difference with the calculated results.

Nowadays, Neutron Transmutation Doping (NTD) of silicon ingots is successfully used to produce the n-type semiconductor. NTD’s main advantage, which is to say its uniform resistivity distribution, leads to semiconductor materials with fewer defects and, accordingly, better performance. Radial uniformity usually is achieved by rotating the ingot during irradiation. However, due to neutron attenuation, the neutron reaction rate in the ingot’s center region generally is lower than in the outer region. Such radial non-uniformity becomes severe when the diameter of the ingot is larger, or when materials with a high-neutron-reaction cross-section are used. In this study, a slit method to selectively shield neutrons that react with the ingot’s outer region and increase radial uniformity thereby was devised and herein is proposed. Dark matter (DM), graphite, stainless steel (SS) and borated stainless steel (BSS) were considered as shielding materials in a Monte Carlo-simulation-based evaluation of the radial uniformity of a Silicon Carbide (SiC) ingot. In the case of DM, the relative reaction rate at the center region of the ingot was increased; graphite did not enhance the radial uniformity; for both SS and BSS, the radial uniformity was under 1% with the optimized slot size. Therefore, it is expected that if the NTD technique is used in concert with the proposed slit method, semiconductors of exceedingly uniform dopant distribution can be produced.

To improve the accuracy and safety of irradiation tests in High flux Advanced Neutron Application ReactOr (HANARO), the nuclear energy deposition rate, which is called nuclear heating, was estimated for an irradiation capsule with an iridium sample in the irradiation hole in order. The gamma rays emitted from the radioisotopes (RIs) of the structural materials such as flow tubes of fuel assemblies and heavy water reflector tank were considered as radiation source. Using the ORIGEN2.1 code, emission rates of delayed gamma rays were calculated in consideration of the activation procedure for 8 years and 2 months of HANARO operation. Calculated emission rates were used as a source term of delayed gamma rays in the MCNP6 code. By using the MCNP code, the nuclear heating rates of the irradiation capsules in the inner core, outer core, and heavy water reflector tank were estimated. Calculated nuclear heating in the inner core, outer core, and heavy water reflector tank were 200–260 mW, 80–100 mW, and 10 mW, respectively. Keywords: Delayed Gamma-ray, HANARO, MCNP, Nuclear Heating, Neutron Irradiation, ORIGEN, Reactor Structure

A new capsule design was prepared and tested at the High Flux Advanced Neutron Application Reactor (HANARO) for neutron irradiation of the core materials of research reactors at a low temperature (...

A neutron irradiation control method using a self-powered neutron detector (SPND) and zirconium foils was developed for neutron transmutation doping (NTD) application at HANARO. An SPND was installed at a fixed position of the upper part of the sleeve at the HANARO NTD hole for real-time monitoring of the neutron irradiation. The SPND signal output was significantly affected by the in-core conditions and surroundings of the facility. Furthermore, the SPND signal varied by $\sim 15\%$ throughout the reactor cycle according to the change of the control rod position. However, it was also confirmed that the variation of the neutron flux on the silicon ingots inside the irradiation rig was not significant for the control rod motion. In this procedure, the process for determining the thermal neutron flux by using zirconium foils was established for achieving efficient NTD irradiation. Accordingly, the relationship between the ratio of the neutron flux to the SPND output and the control rod position was deduced. Silicon irradiation experiments for transmutation doping were performed using the established relationship. The irradiated neutron fluence was controlled within ±1.3% of the target value. The mean irradiation/target ratio of the fluence was 0.9992, and the standard deviation was 0.0071. Thus, we confirmed that the produced irradiation was extremely accurate. This procedure can be useful in SPND applications requiring real-time high-accuracy neutron fluence measurements.

A calibration technology of the self-powered neutron detectors (SPNDs) using a large vertical irradiation hole of HANARO is developed. The 40 Rh-SPNDs are installed on the polycarbonate plastic support, and the gold wires with the same length as the effective length of the rhodium emitter of the SPND are also installed to measure the neutron flux on the SPND. They are irradiated at a low reactor power, and the SPND current is measured using the pico-ammeter. The external gamma-rays which affect the SPND current response are analyzed using the Monte Carlo simulation for various irradiation conditions in HANARO. It is confirmed that the effect of the external gamma-rays to the SPND current is dependent on the reactor characteristics, and that it is affected by materials around the detector. The current signals due to the external gamma-rays can be either positive or negative, in that the net flow of the current may be either in the same or the opposite direction as the neutron-induced current by the rhodium emitter. From the above procedure, the effective calibration methodology of multiple SPNDs using the large hole of HANARO is developed. It could be useful for the calibration experiment of the neutron detectors in the research reactors.

As HANARO has been recently required to support new R&D relevant to future nuclear systems requiring much higher neutron fluence, two types of bottom rod tip of the capsule were preliminarily prepared. The first one is a conventional design made of STS304 and welded using a tungsten inert gas (TIG) welding method. The other is a new design made of STS316L and welded using electron beam (EB) welding to strengthen the fatigue property of the rod tip. During the out-pile testing, they failed after 40 and 203 days, respectively. The fracture surfaces were examined using microscopes and the maximal applied stresses were estimated. The combination of these stresses was proved to be sufficient to cause a fatigue failure of the rod tip of the capsule. Based on the failure analysis, an optimized design of the rod tip of the capsule was made for long-term irradiation testing. It was designed to improve the welding and fatigue properties, to decrease the applied stress on the rod tip, and to fundamentally eliminate the effect of residual stress due to welding. The newly designed capsule was safely out-pile-tested up to 450 days and will be utilized for HANARO irradiation testing.

The modeling method to establish a depleted-core model for the Monte Carlo simulation was verified by comparing the analysis results with the neutron flux distribution obtained by a zirconium activation method and a reaction rate of 30Si(n, γ) 31Si obtained using a resistivity measurement method. First, the modeling method used to build the depleted-core model using the HANAFMS core management system is explained. Calculation and measurement methods are also described including the reactor condition, irradiation device, and data treatment methodology. Finally, the viability of the proposed modeling method was discussed through a comparison of the calculated and measured results.

We measure the thermal neutron capture cross sections of low abundance stable nuclei. The experiments are done with natural foils with known relative isotopic abundances. The cross sections can be obtained by comparing the yields of ’s from the neutron captures by various isotopes. The thermal neutron capture cross sections for isotopes with high abundances are already well measured, and we can obtain the cross sections for nuclei with low abundances which are not known or measured with large errors by simply getting the ratios of gamma yields. For the cross sections of the 152 Gd(n; ) reaction at thermal neutron energy, there are only two old measurements, 1100 100 barn and 735 10 barn, and there is a big discrepancy in the two values. We have measured the neutron capture cross section of the 152 Gd with a natural gadolinium foil. The cross section was obtained by comparing the yields of ’s from the neutron captures by 158 Gd. We have obtained a new cross section value of 614.7 57.8 barn for the 152 Gd(n; ) reaction.

The activation products of aluminum and argon are key radionuclides for operational and environmental radiological safety during the normal operation of open-tank-in-pool type research reactors using aluminum-clad fuels. Their activities measured in the primary coolant and pool surface water of HANARO have been consistent. We estimated their sources from the measured activities and then compared these values with their production rates obtained by a core calculation. For each aluminum activation product, an equivalent aluminum thickness (EAT) in which its production rate is identical to its release rate into the coolant is determined. For the argon activation calculation, the saturated argon concentration in the water at the temperature of the pool surface is assumed. The EATs are 5680, 266 and 1.2 nm, respectively, for Na-24, Mg-27 and Al-28, which are much larger than the flight lengths of the respective recoil nuclides. These values coincide with the water solubility levels and with the half-lives. The EAT for Na-24 is similar to the average oxide layer thickness (OLT) of fuel cladding as well; hence, the majority of them in the oxide layer may be released to the coolant. However, while the average OLT clearly increases with the fuel burn-up during an operation cycle, its effect on the pool-top radiation is not distinguishable. The source of Ar-41 is in good agreement with the calculated reaction rate of Ar-40 dissolved in the coolant.

Recently, the neutron irradiation for large diameter silicon (Si)-ingots of more than 8" diameter is requested to satisfy the demand for the neutron transmutation doping silicon (NTD-Si). By increasing the Si-ingot diameter, the radial non-uniformity becomes larger due to the neutron attenuation effect, which results in a limit of the feasible diameter of the Si-ingot. The current evaluation method has a certain limit to precisely evaluate the radial uniformity of Si-ingot because the current evaluation method does not consider the effect of the Si-ingot diameter on the radial uniformity. The objective of this study is to propose a new evaluation method of radial uniformity by improving the conventional evaluation approach. To precisely predict the radial uniformity of a Si-ingot with large diameter, numerical verification is conducted through comparison with the measured data and introducing the new evaluation method. A new concept of a gradient is introduced as an alternative approach of radial uniformity evaluation instead of the radial resistivity gradient (RRG) interpretation. Using the new concept of gradient, the normalized reaction rate gradient (NRG) and the surface normalized reaction rate gradient (SNRG) are described. By introducing NRG, the radial uniformity can be evaluated with one certain standard regardless of the ingot diameter and irradiation condition. Furthermore, by introducing SNRG, the uniformity on the Si-ingot surface, which is ignored by RRG and NRG, can be evaluated successfully. Finally, the radial uniformity flattening methods are installed by the stainless steel thermal neutron filter and additional Si-pipe to reduce SNRG.

Due to the national demand for a cold neutron beam utilization, a cold neutron research facility project has been carried out since July 2003 to install a cold neutron source (CNS) in HANARO. The CNS adopts a two-phase thermo-siphon of liquid hydrogen as a working fluid to remove a heat load. The CNS consists of an in-pool assembly (IPA) and other components including a hydrogen buffer tank. The liquid level in the moderator cell is measured by a gamma-ray densitometer under a normal cold operation, and the cool down time to fill the required liquid hydrogen into the moderator cell is numerically simulated and compared with the experimental data from a thermo-siphon mock-up test. The self-regulating characteristic of the two-phase thermo-siphon loop is also confirmed by a comparison of a model simulation’s results, which use a thermodynamic model, with the experiment’s results. From these results, the HANARO-CNS is found to maintain the required conditions for a stable cold neutron production against a heat load variation along with the reactor’s power. Furthermore, the thermodynamic behavior in the IPA is observed to determine whether or not the integrity of the IPA is maintained under an abnormal condition of the helium refrigeration system.

The neutron transmutation doping (NTD) method was applied to the initially p-type silicon in order to extend the NTD applications at HANARO. The relationship between the irradiation neutron fluence and the final resistivity of the initially p-type silicon material was investigated. The proportional constant between the neutron fluence and the resistivity was determined to be 2.3473x10(19)nOmegacm(-1). The deviation of the final resistivity from the target for almost all the irradiation results of the initially p-type silicon ingots was at a range from -5% to 2%. In addition, the burn-up effect of the boron impurities, the residual (32)P activity and the effect of the compensation characteristics for the initially p-type silicon were studied. Conclusively, the practical methodology to perform the neutron transmutation doping of the initially p-type silicon ingot was established.

The void fraction in the liquid hydrogen used for the moderator of the HANARO cold neutron source (CNS) was measured by using a gamma densitometer technique. A mock-up of the HANARO CNS facility with an electric heating system as the heat source instead of radiations was constructed. The photon transmissions through the hydrogen moderator were simulated to search for an optimum experimental condition. From the simulation, it was confirmed that Am-241 was suitable for the measurement of the void fraction in the liquid hydrogen medium. A gamma densitometer using the Am-241 gamma-ray source was designed and installed at the mock-up of the CNS. The attenuation of 59.5 keV gamma-rays from the Am-241 through the hydrogen medium was measured by using an HPGe detector. The void fraction was determined using the amount of the gamma-ray attenuation. The void fractions in the hydrogen moderator were measured for stable thermo-siphon loops with several electric heat loads applied to the moderator cell of the CNS mock-up. The longitudinal distribution of the void fraction inside the moderator cell was also determined. The void fraction measured at a heat load of 720 W had values of 8-41% depending on the height from the bottom of the moderator cell. The overall void fraction was obtained by volume-weighted averaging of its longitudinal distribution. The void fraction at the nuclear heating power expected at the normal operation condition of the HANARO CNS facility was determined to be about 20%. The large uncertainty was expected in the void fraction determination by a gamma densitometer for the liquid hydrogen medium with the void fraction less than 10%. When the void fraction of the liquid hydrogen was near 20%, the uncertainty in the void fraction determination by using a gamma densitometer became relatively small, and it was regarded as an acceptable level. The measurements for the void fraction will be very useful for the design and operation of the HANARO CNS.

The gasification rates of several radionuclides dissolved in HANARO pool water were investigated in order to produce the basic data related to the potential release of radioactivity from the reactor pool water to the atmospheric environment. The primary coolant water was collected during 30 MW operation of HANARO. The variation of activity concentrations of several radionuclides in the water with the cooling time were continually measured using the HPGe detection system. The nuclides of interest were Ar-41, Xe-133 and iodine nuclides. The effective half-life of Ar-41 in the water sample was measured to be 0.65±0.02 hours, and its elimination half-life due to the gasification and surface exchange coefficient were determined to be 1.01 hours and 6.89x10−4 cm s−1, respectively. For Xe-133, the elimination half-life due to the gasification and surface exchange coefficient were determined to be 1.44 days and 2.27×10−5 cm s−1, respectively. For iodine nuclides, the measured effective half-lives were similar to ...

The sensitivity of the fuel failure detection system based on the delayed neutron measurement in the primary cooling circuit of a research reactor, HANARO is investigated. The neutrons around the primary cooling pipe during normal operation of HANARO are measured with BF 3 detector, and their count rate is 900 cps. They are regarded as photoneutrons due to the high energy gamma-rays from N-16 and delayed neutrons from the fission of the uranium contaminated on the fuel surface. The contribution of each neutron source is analyzed by measuring the changes of the neutron counts before and after the abrupt shutdown of reactor. In order to estimate the sensitivity of the fuel failure detection, the neutron count rate of BF 3 detector is predicted by Monte Carlo calculation. The generation, transportation and detection of the photoneutrons and the delayed neutrons are simulated for the geometry similar to the experiments. From the calculations and experiments, it is ascertained that the photoneutron contribution to the total count rate is about 20–30%, and that the delayed neutron count rate is expected to about 720 cps. The fission rate in the flow tube of the reactor core by the surface contamination is obtained from the deduced delayed neutron count rate, and it is estimated to 1.66 × 10 5 fissions/cm 3 s. From the MCNP calculation, it is confirmed that this fission rate can originate from the contaminated uranium of 120 μg, which is about 13% of the maximum allowable surface contamination on the fuel surface. The sensitivity of U-235 mass detection by the delayed neutron measurement can be concluded to about 0.2 μg-U 235 /cps. Thus, it is confirmed that the delayed neutron detection is sensitive enough to monitor the fuel failure, and that the neutron count rate is high enough for stable signal with short counting time.

II–VI/III–V heterostructures based on wide-energy band-gap II–VI compound semiconductors have been very attractive because of their potential applications in short-wavelength optoelectronic devices [1–5]. Among these II–VI/III–V mixed heterostructures, ZnSe/GaAs heterostructures have been particularly interesting due to their promising applications for the fabrication of laser diodes and light-emitting diodes operating in the blue-green region of the spectrum [6–10]. However, relatively little work has been performed on II–VI/III–V heterostructures in comparison with III–V/III–V heterostructures due to cross-doping problems resulting from interdiffusion or intermixing during growth and fabrication [11]. Since thermal treatment is necessary for the fabrication processes of several kinds of optoelectronic devices utilizing ZnSe/GaAs heterostructures, the role of the thermal annealing processes is very important in achieving high-performance devices [12]. Therefore, studies of the annealing effects on the surface, the structural, and the optical properties play a very important role in enhancing device efficiency. This letter reports the effect of annealing of ptype ZnSe epilayers grown on n-type GaAs (100) substrates by using molecular beam epitaxy (MBE). Atomic force microscopy (AFM), photoluminescence (PL), and transmission electron microscopy (TEM) measurements were performed in order to investigate the surface, the optical, and the structural properties of as-grown and annealed ZnSe/GaAs heterostructures. The samples used in this study were p-type ZnSe epitaxial layers grown on n-type GaAs (100) substrates in a Riber 32 P system by using MBE. Elemental Zn and Se with purities of 99.9999% were used as the source materials, and the flux ratio, Zn/Se, was approximately 0.5 during ZnSe epilayer growth. The p-type doping was carried out using an rf plasma source to excite nitrogen gas at a pressure of 10−8 Torr in the

The Neutron Transmutation Doping (NTD) of silicon is a method to produce a high quality n-type semiconductor. The NTD technology makes it possible to dope a silicon ingot with an extremely uniform dopant distribution. Korea Atomic Energy Research Institute (KAERI) has been providing the NTD service for 5 and 6 inch silicon ingots from 2003 and 2005 respectively at using HANARO. Coping with recent market demands for a silicon semiconductor by NTD, an additional irradiation facility which has a potential for the 8 inch silicon ingots was developed and the test irradiation is under way from the end of 2008. It is expected that a full scale production of 8 inch NTD-Si will be started from early 2009 and then the total capacity will be increased to 50 tons per year. This paper describes the general characteristics and operation of the NTD facility in HANARO.Copyright © 2009 by ASME

The HANARO ex-core neutron irradiation facility was developed for various applications in the boron neutron capture therapy (BNCT) field, and its characteristics have been investigated. In order to obtain a sufficient thermal neutron flux with a low level contamination of fast neutrons and gamma-rays, a radiation filtering method is adopted. The radiation filter has been designed by using a silicon single crystal cooled by liquid nitrogen and a bismuth crystal. The installation of the main components of the irradiation facility and the irradiation room are finished. Experimental measurements of the neutron beam characteristics have been performed by using bare and cadmium covered gold foils and wires. The in-phantom neutron flux distribution was measured for a flux mapping inside the phantom. The gamma-ray dose was determined by using TLD-700 thermoluminescence dosimeters. The thermal and fast neutron fluxes and the gamma-ray dose were calculated by using the MCNP code, and they were compared with experimental data. The thermal neutron flux and Cd ratio which can be obtained at this facility are 1.49E9 n/cm2s and 152, respectively. The maximum neutron flux inside the phantom was measured to be 2.79E9 n/cm2s at a depth of 3 mm in the phantom. The two-dimensional in-phantom neutron flux distribution was determined, and the significant neutron irradiation was observed within 20 mm from the phantom surface. The gamma-ray dose rate for the free beam condition is expected to be about 80 cGy/hr. These experimental results are reasonably well supported by the calculated values of the facility design code. This HANARO thermal neutron facility can be used not only for a clinical trial but also for various basic irradiation researches of the BNCT field., 15 pages including 10 figures and 5 tables