• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.2 no.1
• /
• pp.1-11
• /
• 2004

The hot cell facility for research activities related to the lithium reduction of spent fuel, which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations peformed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}, 2.97{\times}10^{-3} and 1.01{\times}10{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}, 2.99{\times}10^{-3} and 7.88{\times}10^{-2}$ mSv/h, respectively, The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources. Therefore, it is confirmed that the radiological design for hot cell facility satisfies the Korean criterion of 0.01 mSv/h for the operation area and 0.15 mSv/h for the service (maintenance) area.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.14 no.4
• /
• pp.228-236
• /
• 1995

In this paper the method to evaluate shielding deficiency by gamma scanning test was presented and verified theoretically by Monte Carlo code which is one of the best effective method for radiation shielding calculation. The cylindrical shielding model was selected to evaluate shielding deficiency by gamma scanning test. First, the reference shielding according to the design requirement of cask was fabricated specially and reference values were measured with Co-60 source and scintillation detector. As a result with which calculated the reference values, it is shown that maximum deficiency thickness for lead of true cylindrical shielding model was 12mm. To verify this, thickness of lead was calculated by MCNP code and maximum deficiency thickness was 11.6mm. The experimental result obtained by the use of reference shielding was in good agreement with the theoretical result within 4.1%. So, this method can be applied to inspect the shielding ability for great shielding or cask which the radioactive material is used. To perform measurement more exactly, the further work on the development of measuring equipment to display the results on the screen will be required.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.06a
• /
• pp.457-463
• /
• 2005

KHNP(Korea Hydro and Nuclear Power Ltd., Co.) is developing a HIC transport package which is satisfying domestic and IAEA regulations and NETEC(Nuclear Environment Technology Institute) is conducting a conceptual design. In this study, the shielding thickness was calculated using the data from radionuclide assay program which is currently using in nuclear sites and Micro Shield code. Considering the structural safety, carbon steel was chosen as shielding material and the shielding thickness was calculated for 500 R/hr and 100 R/hr at HIC surface, respectively. Through the shielding analysis, it was evaluated that the regulation limit is satisfied when the shielding thickness is 22 cm for 500 R/hr and 17 cm for 100/hr.

• Journal of Radiation Protection and Research
• /
• v.3 no.1
• /
• pp.29-32
• /
• 1978

A shielding calculation has been carried out for a storage vault of $5292(42{\times}42{\times}3)$ waste drums in which the mixed radioactive gamma-emitters are contained. The required ordinary concrete shielding thickness seems to be approximately 50cm. The results in terms of dose rate for polyenergy gammas appear to be considerably higher than those of the averaged energy gamma.

• Journal of radiological science and technology
• /
• v.39 no.4
• /
• pp.595-600
• /
• 2016

The diagnostic multileaf collimator(MLC) was designed for patient dose reduction in diagnostic radiography We used monte carlo simulation code (MCNPX, LANL, USA) to evaluate efficiency of shielding material for making diagnostic MLC as preliminary study. The diagnostic radiography unit was designed using SRS-78 program according to tube voltage (80,100,120 kVp) and acquired energy spectrums. The shielding material was SKD11 alloy tool steel that is composed of 1.6% carbon(C), 0.4% silicon(Si), 0.6% manganese (Mn), 5% chromium (Cr), 1% molybdenum(Mo) and vanadium(V). The density of it was $7.89g/cm^3$.Using tally card 6, we calculated the shielding efficiency of MLC according to tube voltage. The results was that 98.3% (80 kVp), 95.7 %(100 kVp), 93.6% (120 kVp). We certified efficiency of diagnostic MLC fabricated from SKD11 alloy steel by monte calro simulation. Based on the results, we designed the diagnostic MLC and will develop the diagnostic MLC for reduction of patient dose in diagnostic radiography.

• Journal of Radiation Protection and Research
• /
• v.19 no.1
• /
• pp.13-22
• /
• 1994

The MCNP 4.2 code was used to calculate the thermal neutron flux distributions for $(n,\;{\gamma})$reaction in mainshell, annular plate, and subshell of the calandria of a CANDU 6 plant during operation. The thermal neutron flux distributions in calandria mainshell, annular plate, and subshell were in the range of $10^{11}{\sim}10^{13}\;neutrons/cm^2-sec$ which is somewhat higher than the previous estimates calculated by DOT 4.2 code. As an application to shielding analysis, photon dose rates outside the side and bottom shields were calculated. The resulting dose rates at the reactor accessible areas were below design target, $6 {\mu}Sv/h$. The methodology used in this study to evaluate the thermal neutron flux distribution for $(n,\;{\gamma})reaction$ can be applied to radiation shielding analysis of CANDU 6 type plants.

• Journal of Radiation Protection and Research
• /
• v.31 no.4
• /
• pp.203-210
• /
• 2006

Purpose : To estimate the dose to the embryo/fetus of a pregnant patient with brain tumors, and to design an shielding device to keep the embryo/fetus dose under acceptable levels Materials and Methods : A shielding wall with the dimension of 1.55 m height, 0.9 m width, and 30 m thickness is fabricated with 4 trolleys under the wall. It is placed between a Patient and the treatment head of a linear accelerator to attenuate the leakage radiation effectively from the treatment head, and is placed 1 cm below the lower margin of the treatment field in order to minimize the dose to a patient from the treatment head. An anti-patient scattering neck supporters with 2 cm thick Cerrobend metal is designed to minimize the scattered radiation from the treatment fields, and it is divided into 2 section. They are installed around the patient neck by attach from right and left sides. A shielding bridge for anti-room scattered radiation is utilized to place 2 sheets of 3 mm lead plates above the abdomen to setup three detectors under the lead sheets. Humanoid phantom is irradiated with the same treatment parameters, and with and without shielding devices using TLD, and ionization chambers with and without a build-up cap. Results : The dose to the embryo/fetus without shielding was 3.20, 3.21, 1.44, 0.90 cGy at off-field distances of 30, 40, 50, and 60 cm. With shielding, the dose to embryo/fetus was reduced to 0.88, 0.60, 0.35, 0.25 cGy, and the ratio of the shielding effect varied from 70% to 80%. TLD results were 1.8, 1.2, 0.8, 1.2, and 0.8 cGy. The dose measured by the survey meter was 10.9 mR/h at the patient's surface of abdomen. The dose to the embryo/fetus was estimated to be about 1 cGy during the entire treatment. Conclusion : According to the AAPM Report No 50 regarding the dose limit of the embryo/fetus during the pregnancy, the dose to the embryo/fetus with little risk is less than 5 cGy. Our measurements satisfy the recommended values. Our shielding technique was proven to be acceptable.

• The Journal of Korean Society for Radiation Therapy
• /
• v.13 no.1
• /
• pp.122-125
• /
• 2001

I. 목적 : Primus의 Multi Leaf Collimator는 X측이 29쌍으로 이루어 져 있으며 그중27쌍의 leaf은 1cm이고 2쌍은 6.5cm의 leaf으로 구성되어있다. 이러한 이유로 Paraaortic node를 포함하는 자궁경부암 치료시 현재는 Y가 27cm 이상인 경우는 차폐블럭을 제작하여 치료를 시행하나 차폐 블럭의 제작에 따른 업무의 지연과 차폐 블럭이 무거워져 치료시 환자에게 떨어질 위험을 제거 하기 위해 Asymmetric Field로 Multi Leaf Collimator를 사용한 결과를 보고하고자 한다. II. 재료 및 방법 : 모의 치료(Ximatron, Varian, USA)시 $Y_1$측을 15cm을 기준으로 하여 $Y_2$측을 변경하면서 Field size를 결정한다. ($Y_2$측은 20cm가 최대, 즉 Y측은 35cm까지 적용) 이러한 방법으로 Multi Leaf Collimator를 사용한 환자와 기존의 차폐블럭을 제작하여 치료한 환자와의 업무 개선사항을 확인하기 위하여 실제 공작실 업무 담당자의 블럭 제작 시간과 Beam Shaper를 이 용해 Multi Leaf Collimator를 입력하는 시간을 상호 비교하여 단축된 시간을 조사하였다. III. 결과 : 차폐블럭을 대신해 Multi Leaf Collimator를 이용함으로써 치료실에서 환자에 대한 위험요소(차폐블럭이 무겁다)를 사전에 제거 할 수 있었고 공작실에서 블럭 제작 시간과 LANTIS를 이용해 MLC를 입력하는 시간을 실측 한 결과업무의 시간이 120분에서 5분으로 단축되는 효과가 있었다. 전산화 계획 실에서 선량 계산시 OAR Factor값을 고려하여야 한다. IV. 결론 : Paraaortic node를 포함한 자궁경부암 환자의 차폐부위는 모양이 거의 일직선이기 때문에 Mu]ti Leaf Collimator를 사용하기에 용이 한 치료 부위이다. 하지만 큰 Field size로 인한 불편함이 있었다. 이러한 제 약성을 Asymmetric Field를 이용해서 Multi Leaf Collimator의 사용을 가능하게 하고 차폐 블록의 제작과정과 치료 시에 발생되는 근무자의 업무의 손실을 줄이고 환자에 대한 위험성 을 해결하였다.

• Journal of radiological science and technology
• /
• v.26 no.3
• /
• pp.51-57
• /
• 2003

Mitigation of fast neutron irradiation damage on reactor vessel and improvement of mechanical integrity are desired for the successful plant life-time extension. In this study, the performance of metallic hydride for this application is reviewed and compared. First, selected prospective metallic hydrides are evaluated by MCNP code and put into the attenuation test using Cf-252 neutron source. Since for the reactor application high moderation and reflection with no absorption are favored, Z factor is introduced for the evaluation. According to the Z value estimation $ZrD_2$ and $TiD_2$ are turned out to be the most favorable fast neutron shielding materials. More thorough evaluation by computer simulation and experimentally, will be followed.

• Journal of Radiation Protection and Research
• /
• v.39 no.1
• /
• pp.21-29
• /
• 2014

A new cold neutron triple-axis spectrometer (Cold-TAS) was recently constructed at the 30 MWth research reactor, HANARO. The spectrometer, which is composed of neutron optical components and radiation shield, required a redesign of the segmented monochromator shield due to the lack of adequate support of its weight. To shed some weight, lowering the height of the segmented shield was suggested while adding more radiation shield to the top cover of the monochromator chamber. To investigate the radiological effect of such change, we performed MCNPX simulations of a few different configurations of the Cold-TAS monochromator shield and obtained neutron and photon intensities at 5 reference points just outside the shield. Reducing the 35% of the height of the segmented shield and locating lead 10 cm from the bottom of the top cover made of polyethylene was shown to perform just as well as the original configuration as radiation shield excepting gamma flux at two points. Using gamma map by MCNPX, it was checked that is distribution of gamma. Increased flux had direction to the top and it had longer distance from top of segmented shield. However, because of reducing the 35% of the height, height of dissipated gamma was lower than original geometry. Reducing the 35% of the height of the segmented shield and locating lead 10cm from the bottom of the top cover was selected. After changing geometry, radiation dose was measured by TLD for confirming tester's safety at any condition. Neutron(0.21 ${\mu}Svhr^{-1}$) and gamma(3.69 ${\mu}Svhr^{-1}$) radiation dose were satisfied standard(6.25 ${\mu}Svhr^{-1}$).