• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.219-222
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• 2013

Alloy 690 has been selected as a steam generator tubing material for SMART owing to a near immunity to primary water stress corrosion cracking. The steam generators of SMART are faced with a neutron flux due to the integrated arrangement inside a reactor vessel, and thus it is important to know the irradiation effects of the thermal conductivity of Alloy 690. Alloy 690 was irradiated at HANARO to fluences of (0.7-28) ${\times}10^{19}n/cm^2$ (E>0.1MeV) at $250^{\circ}C$, and its thermal conductivity was measured using the laser-flash equipment in the IMEF. The thermal conductivity of Alloy 690 was dependent on temperature, and it was a good fit to the Smith-Palmer equation, which modified the Wiedemann-Franz law. The irradiation at $250^{\circ}C$ did not degrade the thermal conductivity of Alloy 690, and even showed a small increase (1%) at fluences of (0.7~28) ${\times}10^{19}n/cm^2$ (E>0.1MeV).

• Journal of Radiation Industry
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• v.6 no.4
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• pp.317-322
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• 2012

In this study, the preparation of a temperature-responsive poly(N-isopropylacrylamide) (PNIPAAm)-grafted surface was performed using an eco-friendly and biocompatible ion beam-induced surface graft polymerization. The surface of a perfluoroalkoxy (PFA) film was activated by ion implantation and N-isopropylacrylamide (NIPAAm) was then graft polymerized selectively onto the activated regions of the PFA surfaces. Based on the results of the peroxide concentration and grafting degree measurements, the amount of the peroxide groups formed on the implanted surface was dependant on the fluence, which affected the grafting degree. The results of the FT-IR-ATR, XPS, and SEM confirmed that the NIPAAm was successfully grafted onto the implanted PFA. Moreover, the contact angle measurement at different temperatures revealed that the surface of the PNIPAAm-grafted PFA film was temperature-responsive.

• Corrosion Science and Technology
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• v.20 no.4
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• pp.189-195
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• 2021

A technology for designing and licensing a dedicated radiation shielding facility needs to be developed for safe and efficient operation an R&D center. Technology development is important for smooth operation of such facilities. Causes of damage to internal structures (such as baffle former bolt (BFB) of pressurized water reactor) of a nuclear power reactor should be analyzed along with prevention and countermeasures for similar cases of other plants. It is important to develop technologies that can comprehensively analyze various characteristics of internal structures of long term operated reactors. In high-temperature, high-pressure operating environment of nuclear power plants, cases of BFB cracks caused by irradiated assisted stress corrosion cracks (IASCC) have been reported overseas. The integrity of a reactor's internal structure has emerged as an important issue. Identifying the cause of the defect is requested by the Korean regulatory agency. It is also important to secure a foundation for testing technology to demonstrate the operating environment for medium-level irradiated testing materials. The demonstration testing facility can be used for research on material utilization of the plant, which might have highest fluence on the internal structure of a reactor globally.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1472-1482
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• 2017

Zirconium alloy cladding tube specimens were irradiated at $380^{\circ}C$ up to a fast neutron fluence of $7.5{\times}10^{24}n/m^2$ in a research reactor to investigate the effect of neutron irradiation on hydride reorientation and mechanical property degradation. Cool-down tests from $400^{\circ}C$ to $200^{\circ}C$ under 150 MPa tensile hoop stress were performed. These tests indicate that the irradiated specimens generated a smaller radial hydride fraction than did the unirradiated specimens and that higher hydrogen content generated a smaller radial hydride fraction. The irradiated specimens of 500 ppm-H showed smaller ultimate tensile strength and plastic strain than those characteristics of the 250 ppm-H specimens. This mechanical property degradation caused by neutron irradiation can be explained by tensile hoop stress-induced microcrack formation on the hydrides in the irradiation-damaged matrix and subsequent microcrack propagation along the hydrides and/or through the matrix.

• Journal of Radiation Protection and Research
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• v.10 no.2
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• pp.89-95
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• 1985

An alternative estimator for dose equivalent was derived. The original LET distribution concept was transformed into a charged particle fluence spectrum concept along with the definition of an average quality factor named slowing-down averaged quality factor by adopting the continuous slowing down approximation. With the alternative estimator, the dose equivalent delivered into a receptor located in a given radiation field can be directly and conveniently estimated in a Monte Carlo procedure. The slowing-down averaged quality factors for the energy range below 10 MeV were evaluated and tabulated for the charged particles which may be generated from the interactions of neutron with the nuclei composing soft tissue.

• Journal of Mechanical Science and Technology
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• v.19 no.6
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• pp.1378-1389
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• 2005

This work provides the fundamental knowledge of energy transport characteristics during very short-pulse laser heating of semiconductors from a microscopic viewpoint. Based on the self-consistent hydrodynamic equations, in-situ interactions between carriers, optical phonons, and acoustic phonons are simulated to figure out energy transport mechanism during ultrafast pulse laser heating of a silicon substrate through the detailed information on the time and spatial evolutions of each temperature for carriers, longitudinal optical (LO) phonons, acoustic phonons. It is found that nonequilibrium between LO phonons and acoustic phonons should be considered for ultrafast pulse laser heating problem, two-peak structures become apparently present for the subpicosecond pulses because of the Auger heating. A substantial increase in carrier temperature is observed for lasers with a few picosecond pulse duration, whereas the temperature rise of acoustic and phonon temperatures is relatively small with decreasing laser pulse widths. A slight lagging behavior is observed due to the differences in relaxation times and heat capacities between two different phonons. Moreover, the laser fluence has a significant effect on the decaying rate of the Auger recombination.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.71-79
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• 1993

Form the pure Maxwellian distribution(kT= 1.42MeV), the effects upon calibration factors of encapsulating a $^{252}Cf$ spontaneous fission neutron source were investigated to establish a standard neutron field in the Secondary Standard Dosimetry Laboratory at Korea Atomic Energy Research Institute(KAERI). A Monte Carlo code MCNP was used in simulating the encapsulation SR-Cf-100 and SR-Cf-1273 to be real conditions. The anisotropy(FI) and fluence-to-dose equivalents conversion factors$(H/{\Phi})$ were evaluated and compared with other results. As the results, the FI was determined to be 1.061 at ${\theta}=90^{\circ}$ with ${\pm}0.2%$ statistical error and the $(H/{\Phi})$ was evaluated to be $333.9 [pSv\;cm^2]\;with\;{\pm}0.5%$ statistical error, which is lower by 1.8% than that recommended by the ISO 8529. This means physically that the neutron spectrum of the unmoderated $^{252}Cf$ source in KAERI is a little more softened than that by the ISO.

• Korean Journal of Materials Research
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• v.9 no.3
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• pp.327-332
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• 1999

The recovery activation energy, the order of reaction and the recovery rate constant were detemined by isochronal and isothermal annealing treatment to investigate the recovery behaviors of neutron irradiated Mn-Mo-Ni low alloy steels$(fluence: 2.3\times10^{19}ncm^{-2}, 553K, E\geq1.0 MeV)$. Vickers microhardness tests were conducted to trace the recovery behavior after heat treatments. The results were analyzed in terms of recovery stages, behavior of responsible defects and recovery kinetics. It was shown that recovery occurred through two annealing stages(stage I : 703-753K, stage n : 813-873K) with re$\infty$very activation energies of 2.5 eV and 2.93 eV for each stage I and n, respectively. From the comparison of unirradiated and irradiated isochronal anneal curves, a radiation anneal hardening(RAH) peak was identified at around 813K. Most of recovery have occurred during about 120 min irrespective of isothermal annealing temperatures of 743K and 833K. Recovery rate constants were determined to be $3.4\times10^{-4}min^{-1} and 7.1\times10^{-4}min^{-1}$ for stage I and II, respectively. The order of reaction was about 2 for both recovery stages. Comparing the obtained data with those of previously reported results on neutron irradiated Mn- Mo- Ni steels, the thermal recovery be­havior of the present material seems to occur by the dissociation of point defect clusters formed during irradiation, and by the recombination process of self-interstitials and vacancies from dissociated vacancy clusters.

• Journal of Radiation Protection and Research
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• v.7 no.1
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• pp.34-48
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• 1982

The prime purpose of this study is to realize an index quantity, absorbed dose index, defined by the ICRU for the characterization of ambient radiation level at any location for the purpose of radiation protection. The experiment has been designed to be carried out in two phases, namely, preliminary and main experiment. In the primary study a 30cm diameter sphere of polyethylene was used, while in the main experiment that of tissue equivalent material was fabricated and used. Both experiments were performed in the gamma-ray fields of $^{137}Cs\;and\;^{60}Co$, and in a neutron beam of thermal column of the TRIGA MARK-II research reactor. In the measurement of gamma-ray absorbed dose TLD-700 $(^{7}LiF)$ chips were used, and for the neutron dose both Au activation foils and TLD chips (TLD-600 $(^{6}LiF)$ and TLD-700 for the discrimination of gamma-ray contribution) were used. Theoretical assessment of the absorbed dose in the sphere phantom has been carried out in accordance with the Ehrlich's idea that deduced on the basis of Burlin's cavity theory in the case of gamma-ray irradiation. For the analysis of neutron dose fluence-KERMA rate conversion method was used. The explanation on the dose assessment is given in detail. Results obtained were numerically and statistically analyzed and the depth dose distributions are presented in the graphic forms with normalized values. In the concluding remarks, the possibility and difficulty of realizing the index quantity, including questions and problems to be solved are mentioned.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2607-2612
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• 2020

The present work was aimed to show the possibility of using hydrogel (acrylic/boric acid) for evaluation of the neutron radiation shielding. The influence of acrylic acid concentration, different gamma doses and relative contents of boric acid were studied. The physical properties and the thermomechanical stability of the studied samples were investigated. The shielding property of the composite for neutron was tested by Pu-Be neutron source (5 Ci) under room temperature. The neutron fluence rates and gamma fluxes were measured using a stilbene organic scintillator. The macroscopic effective removal cross-section Σ_R (cm^-1) of fast neutrons and total attenuation coefficient μ (cm^-1) of gamma rays has been studied experimentally. The transmission parameters, the relaxation length (??) and the half-value layer (HVL) were obtained. The obtained results indicated that the addition of boric acid to acrylic acid tends to increase the macroscopic effective removal cross-section Σ_R (cm^-1) to 0.141 compared to 0.094 of ordinary concrete.