• Journal of Radiation Protection and Research
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• 제22권1호
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• pp.29-33
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• 1997

중성자교정실내에서 $D_2O$ 감속 $^{252}Cf$중성자선원을 사용하여 계측기를 교정할 때는 그 계측기에 대한 교정실산란보정 인자를 미리 결정하여야 한다. 이러한 교정실산란보정인자는 계측기의 종류, 교정거리, 교정실형태에 따라 다르게 결정된다. 본 연구에서는 한국원자력연구소에서 운영하는 2차 표준중성자교정실에서 한가지의 열형광선량계와 2가지의 구형검출기에 대한 교정실산란보정인자를 실험적으로 결정하였고 본소의 2차 표준중성자교정실조건에 의하여 이론적으로 예측한 값과 비교하였다. 비교한 결과 실험하여 얻어진 상기의 3가지 계측기에 대한 교정실산란보정인자가 이론적으로 예측한 결과와 최대 약 10% 이내에서 일치하였다.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.939-944
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• 2015

The calibration methods of neutron-measuring devices such as the neutron survey meter have advantages and disadvantages. To compare the calibration factors obtained by the shadow cone method and semi-empirical method, 10 neutron survey meters of five different types were used in this study. This experiment was performed at the Korea Atomic Energy Research Institute (KAERI; Daejeon, South Korea), and the calibration neutron fields were constructed using a $^{252}Californium$ ($^{252}Cf$) neutron source, which was positioned in the center of the neutron irradiation room. The neutron spectra of the calibration neutron fields were measured by a europium-activated lithium iodide scintillator in combination with KAERI's Bonner sphere system. When the shadow cone method was used, 10 single moderator-based survey meters exhibited a smaller calibration factor by as much as 3.1-9.3% than that of the semi-empirical method. This finding indicates that neutron survey meters underestimated the scattered neutrons and attenuated neutrons (i.e., the total scatter corrections). This underestimation of the calibration factor was attributed to the fact that single moderator-based survey meters have an under-ambient dose equivalent response in the thermal or thermal-dominant neutron field. As a result, when the shadow cone method is used for a single moderator-based survey meter, an additional correction and the International Organization for Standardization standard 8529-2 for room-scattered neutrons should be considered.

• 한국방사선학회논문지
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• 제2권3호
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• pp.33-35
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• 2008

In the neutron capture experiment and calculation, the neutron absorption and scattering are very important. Especially these effects are conspicuous in the resonance energy region and below the thermal energy region. In the present study, we obtained energy dependent neutron absorption ratios of natural rhodium in energy region from 0.003 to 100 eV by MCNP-4B Code. The coefficients for neutron absorption was calculated for several types of thickness. In the lower energy region, neutron absorption is larger than higher region, because of large capture cross section (1/v). Furthermore it seems very different neutron absorption in the large resonance energy region. These results are very useful to decide the thickness of sample and shielding materials.

• Journal of Radiation Protection and Research
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• 제14권1호
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• pp.66-70
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• 1989

방사성 중성자선원은 일상적 시험에 있어 표준 중성자 방사선장을 형성하는데 적합하다. 방사선 방어상의 목적으로 사용되는 중성자 측정기기의 교정을 위한 기준 방사선이 ISO TC-85에서 제의되었다. 한국표준연구소 방사선연구실에는 ISO TC-85의 추천사항에 준하여 개인용 중성자 선량계를 교정하기 위하여 $^{252}Cf$와 $^{241}Am-Be$ 선원을 이용한 표준조사시설을 설립하였다. 본 연구에서는 중성자 산란과 선원 비등방성에 연관된 교정상의 보정인자들을 실험에 의하여 결정하였다.

• 한국방사선학회논문지
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• 제2권1호
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• pp.31-34
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• 2008

The neutron capture spectrum for the light nuclide was very useful to study the nuclear structure. In the present study, the capture gamma-ray from the 27-keV resonance of $^{19}F(n,g)^{20}F$ reaction were measured with an anti-Compton NaI(Tl) spectrometer and the 3-MV Pelletron accelerator of the Research Laboratory for Nuclear Reactors at the Tokyo institute of technology. A neutron Time-of-Flight method was adopted with a 1.5 ns pulsed neutron source by the $^7Li(p,n)^7Be$ reaction. In the present experiment, a Teflon(($CF_2$)n) sample was used The sample was disk with a diameter of 90mm. The thickness of sample was determined so that reasonable counting rates could be obtained and the correction was not so large for the self-shielding and multiple scattering of neutrons in the sample, and was 5mm. The primary gamma-ray transitions were compared with previous measurement of Kenny.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.642-649
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• 2016

Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANada Deuterium Uranium (CANDU-6) reactor have been evaluated using the Monte Carlo method. For accurate analysis of the parameters, the Doppler broadening rejection correction scheme was implemented in the MCNPX code to account for the thermal motion of the heavy uranium-238 nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted using MCNPX. The FTC value is evaluated for several burnup points including the mid-burnup representing a near-equilibrium core. The Doppler effect has been evaluated using several cross-section libraries such as ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. The PCR value is also evaluated at mid-burnup conditions to characterize the safety features of an equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, we considered a huge number of neutron histories in this work and the standard deviation of the k-infinity values is only 0.5-1 pcm.

• Nuclear Engineering and Technology
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• 제34권2호
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• pp.154-164
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• 2002

Neutron cross section data on Dy-160, Dy-161, Dy-162, Dy-163 and Dy-164 were calculated and evaluated in the energy range of 1 keV to 20 MeV using a spherical optical model, statistical model and pre-equilibrium model. The energy dependent optical model potential parameters were obtained based on the recent experimental data. The width fluctuation correction in Hauser-Feshbach particle decay and the quantum mechanical approach in pre-equilibrium analysis were introduced and gave a better cross section calculation in EMPIRE-II. The total, elastic scattering and threshold reaction cross sections were evaluated and compared with the evaluated files. The model calculated (n, tot), (n, ${\gamma}$) and (n, p) cross sections were in good agreement with the experimental data in the measured energy range. The results will be applied to ENDF/B-VI for data improvement.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.791-803
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• 2015

The effective cross sections (XSs) in the direct whole core calculation code nTRACER are evaluated by the equivalence theory-based resonance-integral-table method using the WIMS-based library as an alternative to the subgroup method. The background XSs, as well as the Dancoff correction factors, were evaluated by the enhanced neutron-current method. A method, with pointwise microscopic XSs on a union-lethargy grid, was used for the generation of resonance-interference factors (RIFs) for mixed resonant absorbers. This method was modified by the intermediate-resonance approximation by replacing the potential XSs for the non-absorbing moderator nuclides with the background XSs and neglecting the resonance-elastic scattering. The resonance-escape probability was implemented to incorporate the energy self-shielding effect in the spectrum. The XSs were improved using the proposed method as compared to the narrow resonance infinite massbased method. The RIFs were improved by 1% in $^{235}U$, 7% in $^{239}Pu$, and >2% in $^{240}Pu$. To account for thermal feedback, a new feature was incorporated with the interpolation of pre-generated RIFs at the multigroup level and the results compared with the conventional resonance-interference model. This method provided adequate results in terms of XSs and k-eff. The results were verified first by the comparison of RIFs with the exact RIFs, and then comparing the XSs with the McCARD calculations for the homogeneous configurations, with burned fuel containing a mixture of resonant nuclides at different burnups and temperatures. The RIFs and XSs for the mixture showed good agreement, which verified the accuracy of the RIF evaluation using the proposed method. The method was then verified by comparing the XSs for the virtual environment for reactor applicationbenchmark pin-cell problem, as well as the heterogeneous pin cell containing burned fuel with McCARD. The method works well for homogeneous, as well as heterogeneous configurations.

• Nuclear Engineering and Technology
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• 제32권4호
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• pp.361-371
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• 2000

The 123Te(p,n)123I, 124Te(p,n)124I and 124Te(p,2n)123I reactions, among the many reaction channels opened, are the major reactions under consideration from a diagnostic purpose because reaction residuals as the gamma emitters are used for most radiophamaceutical applications involving radioiodine. Based on the available experimental data, the absorption cross sections and elastic scattering angular distributions of the proton-induced nuclear reaction on Te isotopes below 60 MeV are calculated using the optical model code APMNK. The transmission coefficients of neutron, proton, deuteron, trition and alpha particles are calculated by CUNF code and are fed into the GNASH code. By adjusting level density parameters and the pair correction values of some reaction channels, as well as the composite nucleus state density constants of the pre-equilibrium model, the production cross sections and energy-angle correlated spectra of the secondary light particles, as well as production cross sections and energy distributions of heavy recoils and gamma rays are calculated by the statistical plus pre-equilibrium model code GNASH. The calculated results are analysed and compared with the experimental data taken from the EXFOR. The optimized global optical model parameters give overall agreement with the experimental data over both the entire energy range and all tellurium isotopes.