• Journal of Radiation Protection and Research
• /
• v.31 no.3
• /
• pp.129-134
• /
• 2006

A TL pellet for a neutron dose measurement (KCT-306) by embedding a $^6Li$-compound into a $CaSO_4:Dy$ phohphor was developed based upon the technical information of KCT-300. The KCT-300 is an another kind of $CaSO_4:Dy$ TL detector shich was developed at KAERI, in which small amounts of $NH_4H_2PO_4$ have been emvedded as a binding material. This paper presented the optimized manufacturing condition of KCT-306 and compared its sensitivity with that of the commercialized neutron TL pellets. $CaSO_4:Dy$ Phosphor with grain size ranging less than $45{\mu}m$ are used for the KCT-306. The optimum $CaSO_4:Dy$ TL phosphor, $^6Li$-compounds and P-compound as the binding material are determined as 20-40wt%, 50-70wt% and 20wt%. The TL pellet combination of our KCT-306/KCT-300, TLD-600/TLD-700 and TLD-600H/TLD-700H(Harshaw) have been irradiated in the neutron/gamma mixed fields from a $D_2O$ moderated $^{252}Cf$ neutron source. The KCT-300, TLD-700 and TLD-700H were used at the same time as gamma ray discriminators in the neutron/gamma mixed fields. It was found that the neutron/gamma response ratios of KCT-306/KCT-300, which were developed in this study, were approximately 4 times higher than those of the commercial TLD-600H/TLD-700H.

• The Journal of Korean Society for Radiation Therapy
• /
• v.17 no.2
• /
• pp.141-145
• /
• 2005

Purpose : It is known that the neutron is generally generated from the photon, its energy is larger than 10 MV. The neutron is leaked in the container inspection system installed at the customs though its energy is below 9 MV. It is needed that the spacial effect of the neutrons released from radiation treatment machine, linac, installed in the medical canter. Materials and Methods : The medical linear accelerator(Clinac 1800, varian, USA) was used in the experiment. Measuring neutron was used bubble detector(Bubble detector, BDPND type, BTI, Canada) which was created bubble by neutron. The bubble detector is located on the medical linear accelerator outskirt in three different distance, 30, 50, 120 cm and upper, lower four point from the iso-center. In addition, for effect on protect material we have measured eight points which are 50 cm distance from iso-center. The SAD(source-axis-distance), distance from photon source to iso-center, is adjusted to 100 cm and the field size is adjusted to $15{\times}15cm^2$. Irradiate 20 MU and calculate the dose rate in mrem/MU by measuring the number of bubble. Results : The neutron is more detected at 5 position in 30, 50 cm, 7 position in 120 cm and with wedge, and 2 position without mount. Conclusion : Though detection position is laid in the same distance in neutron measurement, the different value is shown in measuring results. Also, neutron dose is affected by the additional structure, the different value is obtained in each measurement positions. So, it is needed to measure and evaluate the neutron dose in the whole space considering the effect of the distance, angular distribution and additional structure.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05b
• /
• pp.111-117
• /
• 1998

원자로 압력용기는 운전기간 동안 중성자 조사에 따른 재료의 기계적 성질이 변화되므로, 용기의 건전성 유지여부를 평가하기 위하여 조사시편을 이용한 주기적인 감시시험이 요구된다. 그러나 감시시편은 방사성 물질로서 일반 환경조건에서 시험이 불가능하다. 따라서 국내의 자력기술로 완공된 조사재시험시설의 핫셀 내에서 감시시험의 주요항목인 온도감시자, 충격, 인장, 파괴인성, 그리고 성분분석 등에 대한 시험을 수행하기 위하여 관련된 규정에 합하도록 장비 및 시험 평가기술을 개발하였다.

• Journal of the Korean Chemical Society
• /
• v.18 no.1
• /
• pp.40-46
• /
• 1974

The fission neutron reactions of $^{47}Ti(n.p)^{47}Sc$ and $^{93}Nb(n,{\alpha})^{90m}Y$, along with epicadmium neutron reaction of $^{96}Zr(n,{\gamma})^{97}Zr$ were used for the simultaneous determination of Ti, Nb and Zr in synthetic mixture. Prior to neutron irradiation, Ti, Zr and Nb in the mixture were separated together in one group through the cation exchange column of Dowex $50{\times}8$ resin using 0.5 M ${\alpha}$-hydroxy-iso-butyric acid as the eluent. After irradiation of the eluate the product nuclides, $^{97}Zr$, ^{47}Sc$ and ^{90m}Y$, were eluted sequentially through the same column with 0.5 M ${\alpha}$-HIBA, 0.5 M ${\alpha}$-HIBA-1 N HNO_3 and 0.5 M ${\alpha}$-HIBA-2 N HNO_3$ solution, respectively. The gamma-ray spectrometry was used for the measurement of the gamma-ray activities of the eluted nuclides. The detection limits of Nb, Ti and Zr were found to be 0.2 %, 0.01 % and 0.002 %, respectively.

• Journal of Radiation Protection and Research
• /
• v.23 no.1
• /
• pp.49-57
• /
• 1998

Domestic intercomparison study of the neutron personnel dosemeters was performed for the first time in Korea. Thirteen types of neutron dosemeters from twelve institutions took part in this intercomparison study and the $D_2O$ moderated Cf-252 source of KAERI was used for irradiation. Eight of the fifteen dosemeters submitted by each participant were divided into two groups and each group was irradiated with different doses of the simulated mixed fields of neutron and gamma. The participants assessed their dosemeter reading in terms of the personal dose equivalent, Hp(10), for both neutron and gamma dose. The ratio of the reported dose equivalent to the delivered dose equivalent for comparison between participants ranged from 0.55 to 1.34 for neutron, from 0.54 to 1.32 for gamma and from 0.75 to 1.20 for total dose. This intercomparison results show that all dosemeter processors, especially for neutron category, are able to pass the personnel dosemeter performance test which shall be enforced according to the ordinance of the MOST, No. 96-6.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.06a
• /
• pp.432-432
• /
• 2005

사용후핵연료시료 중에 함유된 요오드(I-127 및 129)를 정량하기 위하여 화학적 방법으로 분리 회수하고 중성자방사화분석법을 이용하였다. 사전실험으로 모의사용후핵연료를 이용하여 회수율을 측정하였다. 모의 및 실제사용후핵연료시료를 $90^{\circ}C$에서 8 M $HNO_3$ 용액으로 용해하고 용해 후 용해용액 중의 잔류 요오드, 응축 및 휘발된 요오드 각각을 정량하였다. 응축 요오드는 핵연료 용해 후 재증류하여 회수하였다. 잔류 및 응축 요오드는 시료의 산화상태를 조절한 후 용매추출로 요오드를 회수한 다음 이온교환 또는 침전법으로 방사화학적으로 분리한 후 중성자방사화분석(RNAA)으로 정량하였다. 제작한 이온교환분리관 및 여과키트에 요오드를 흡착 또는 침전시켜 분리한 다음 중성자조사를 위한 삽입체(Insert)로 이용하였다. 휘발 요오드는 제조한 흡착체(Ag-silica gel)를 담은 흡착관에 포집하고 홉착체를 구간별 균질시료로 만든 다음 비파괴중성자 방사화분석(INAA)으로 정량하였다. 침전 및 흡착 요오드의 화학적 특성을 EPMA(electron probe microanalysis) 분석으로 조사하였다. 요오드 정량결과를 다른 방법으로 비교분석하기 위하여 음이온교환수지상에서 요오드를 정제 및 회수하기 위한 용리거동을 조사하였다.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2006.11a
• /
• pp.87-88
• /
• 2006

사용후 핵연료용 수송용기의 설계 안전평가에서는 이제까지 용기에 수납되는 연료는 미조사, 즉 신연료라 가정해서 보수적으로 임계안전설계를 수행하여 왔다. 이것은 연소에 따른 연료내의 핵연료 물질의 감손 및 생성의 의한 반응도의 변동을 계산 평가하는 것이나 또는 연소로 인해 생성되는 중성자 흡수 핵종의 조성 및 함유량 등을 정확히 계산 평가하는 것이 복잡해서 곤란했던 것으로 그 요인을 들 수 있다. 사용 후 핵연료를 신 연료로 가정하는 등의 불합리성을 해소하고, 안전성을 잃지 않고 사용 후 핵연료 운반용기 들의 경제성을 추구하는 기운이 높아지고, 관련 연구가 적극적으로 진척되게 되었다. 그 결과 연소에 따른 연료내의 핵연료 물질의 감손 생성과 핵분열 생성물 등에 의한 반응도의 저하, 즉 중성자 실효 증배율의 저하를 고려한 것을 사용 후 핵연료용 캐스크 설계 안전평가에 취할 수 있게 되었다. 연소도 크레디트를 채용함으로서 사용후 핵연료내의 핵연료물질량은 실제로 존재하는 양을 사용하는 것이 되므로 초기 농축도가 높은 고연소도 연료에서 그 효과가 보다 크게 될 것이다. 이것은 연소도 크레디트 채용에 따라 연료 바스켓의 중성자흡수제 사용량 감소가 가능해져 사용 캐스크의 수를 줄일 수 있어 경제성 향상이 기대되고 아울러 그이 취급 횟수 및 수송횟수가 감소됨에 따라 안전성의 향상도 기대된다.

• Journal of radiological science and technology
• /
• v.36 no.2
• /
• pp.165-173
• /
• 2013

This study was the estimation of the dose distribution for proton, prompt gamma rays and proton induced neutron particles, in case of exposing the proton beam to polymer gel dosimeter and water phantom. The polymer gel dosimeter was compositeness material of Gelatin, Methacrylic acid, Hydroquinone, Tetrakis and Distilled water. The density of gel dosimeter was $1.04g/cm^3$ which was similar to water. The 72, 116 and 140 MeV proton beams were used in the simulation. Proton beam interacted with the nuclei of the phantom and the nuclei in excited states emitted prompt gamma rays and proton induced neutron particles during the process of de-excitation. The proton particles, prompt gamma rays, proton induced neutron particles were detected by polymer gel dosimeter and water phantom, respectively. The gap of the axis for gel was 2 mm. The Bragg-peak for proton particles in gel dosimeter was similar to water phantom. The dose distribution for proton and prompt gamma rays in gel dosimeter and water phantom was approximately identical in case of 72, 116 and 140 MeV for proton beam. However, in case of proton induced neutron particles for 72, 116 and 140 MeV proton beam, particles were not detected in gel dosimeter, while the Water phantom absorbed neutron particles. Considering the resulting data, gel dosimeter which was developed in the normoxic state attentively detected the dose distribution for proton beam exposure except proton induced neutron particles.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.10a
• /
• pp.103-108
• /
• 1997

압력용기로의 속중성자 조사량 평가를 4군 노달 노심해석코드로 수행하였다. 이 코드는 MCNP에 비해 정확성은 떨어지나, 핵연료 연소의 효과나 핵연료 장전 모형의 영향을 쉽게 고려할 수 있었다. 속중성자 조사량 감소 방안으로서 반사체 차폐 구조물을 설치하는 방안과 노심외곽에 대체 핵연료 집합체를 장전하는 방안을 비교하였다. 신형원전의 경우 가장 효과적인 방안은 물 반사체 영역에 금속 차폐 구조물을 설치하는 것이나 운전중인 원자로의 경우 비록 주기길이의 감소와 핵연료 비용의 증가는 있으나 속중성자 감소 효과에 있어서는 대체 핵연료 집합체의 장전이 대안일 수 있다.

• Analytical Science and Technology
• /
• v.22 no.5
• /
• pp.395-406
• /
• 2009

The determination of iodine in the spent nuclear fuel and the volatile behavior during its acid dissolution have been studied by NAA(neutron activation analysis) and electron probe microanalysis (EPMA). Simulated spent fuels (SIMFUELs) were dissolved in $HNO_3$(1+1) at $90^{\circ}C$ for 8 hours. The iodine remained in a dissolver solution after dissolution, and that condensed in dissolution apparatus and trapped in the adsorbent by volatilization during the dissolution were determined, respectively. The condensed iodine was recovered by the redistillation with $HNO_3$(1+1) after transfer of the dissolver solution. The iodines in the dissolver and redistilled solution were separated by solvent extraction followed by ion exchange or precipitation method and determined by RNAA (radiochemical neutron activation analysis). The ion exchange column and filtration kit used for the isolation of iodine, which were prepared with a polyethylene tube, were used as an insert in the pneumatic tube for neutron irradiation. The iodine volatilized during the dissolution of SIMFUELs was collected in a trapping tube containing Ag-silica gel (Ag-impregnated silica gel) adsorbent, and the distribution of iodine trapped in the adsorbents were determined by EPMA. The adsorbing characteristics shown with the SIMFUELs were compared with those shown with a real spent fuel from the nuclear power plant.