• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.3
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• pp.193-200
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• 2005

[ $^{14}C$ ] and $^{3}H$ in the evaporator bottom (EB) discharged from the Nuclear power plant (NPP) were extracted simultaneously into a gaseous $^{14}CO_{2}$ and liquefied HTO by using the chemical oxidation, which is the method to oxidize samples completely using potassium persulfate and sulfuric acid. The extracted $^{14}C$ and $^{3}H$ were counted by the liquid scintillation counter (LSC) after the quench correction. To examine the recovery of $^{14}C$ using the radioactive standards, $Na_{2}^{14}CO_{3}$, $^{14}C-alcohol$, and $^{14}C-toluene$ as $^{14}C$, and HTO as $^{3}H$ were used. Also, the most suitable method for oxidizing $^{14}C-toluene$, which is difficult to be oxidized, was investigated through FT-IR spectra according to the concentration of sulfuric acid. With the identical method, $^{14}C$ and $^{3}H$ in the EB generated in the NPP were assayed in the range of $8.35{\sim}l.38{\times}10^3$ Bq/g and $2.46{\times}10^2{\sim}1.40{\times}10^4$ Bq/g, respectively.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.48-54
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• 1996

가압경수로의 부분충수 운전중 RHR 계통의 기능상실시 사고완화를 위해 가압기 manway와 증기 발생기 출구공동 manway를 동시에 개방한 경우에 대한 실험결과를 CATHAHR2 코드를 이용하여 해석하였다. 해석을 통해 이 경우에 발생하는 물리적 현상을 이해하고 이와 같은 과도기에 대해 코드 예측능력을 평가하므로 써, 실제 원전에서 사고시 적절한 사고대응 방안을 강구하는데 참고가 될 수 있도록 해석적 근거를 제시하고자 한다. 연구결과 CATHARE2 코드는 실험을 통해 관측된 주요 물리적 현상들을 타당하게 예측하였으나, 가압기와 밀림관의 DP를 과대 예측하여 원자로 상부공동의 최대압력을 실험보다 약 7kPa 높게 예측하였다. 노심 노출시간도 노심에서 기포율 분포를 비현실적으로 예측하여 실험보다 약 500초 지연되었다. 실험과 코드의 모의결과를 통하여 노심 노출은 중력주입에 의한 냉각수 보충만으로 충분히 회복될 수 있음을 확인하였다. CATHARE2 코드는 비록 상세한 현상들에 대해 다소 불확실성을 내포하였으나, 전반적인 거동분석에는 타당한 것으로 판단된다. CATHARE 코드는 노심에서 계면 마찰력을 줄임으로써 노심의 차압을 개선할 수 있었고, guide 튜브의 위치를 고온관 중심선과 일치시켜 guide 튜브내 액체의 hold-up 기간을 개선할 수 있었으며, 가압기의 계면 마찰력을 증가시켜서 밀림관에서 "plug and clearing" 현상을 모의할 수 있었다.모의할 수 있었다.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.278-283
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• 2005

Cement and paraffin waste form were prepared with a acid extraction method for the analysis of radionuclides generated from nuclear power plants. The acid extraction method was carried out with $HNO_3-HCl$ acid. At first, we compared the method with the microwave acid digestion method using SRM. The solutions of decomposed SRM were then analyzed by AAS and ICP-AES. The acid extraction method had shown good results as microwave acid digestion method. This method provided recovery values greater than $80\%$ for metallic elements.

• KIPS Transactions on Computer and Communication Systems
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• v.5 no.9
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• pp.275-284
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• 2016

The software used in the nuclear safety field has been ensured through the development, validation, safety analysis, and quality assurance activities throughout the entire process life cycle from the planning phase to the installation phase. However, this evaluation through the development and validation process needs a lot of time and money, and there are limitations to ensure that the quality is improved enough. Therefore, the effort to calculate the reliability of the software continues for a quantitative evaluation instead of a qualitative evaluation. In this paper, we propose a reliability evaluation method for the software to be used for a specific operation of the digital controller in a nuclear power plant. After injecting weighted faults in the internal space of a developed controller and calculating the ability to detect the injected faults using diagnostic software, we can evaluate the software reliability of a digital controller in a nuclear power plant.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.284-289
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• 2005

According to the nuclear safety regulation policy including the administration of radionuclides in low level radwastes, the evaporator bottoms were mixed with cement to form a stable solidification for identifying the recovery possibility of the C-14. The chemical oxidation method was applied for the extraction of C-14 from the cement waste form. The emitting beta ray of the C-14 extracted from the radwastes was measured with the liquid scintillation counter and calculated by using the quenching correction curves. Only the beta emitting radioactive nuclides of the C-14 in the radwastes was showed the radioactivities with the range of $2.7E+00\;{\sim}\;3.07E+02$ Bq/g.

• Journal of radiological science and technology
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• v.38 no.4
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• pp.337-345
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• 2015

After the Fukushima nuclear accident in 2011, concerns about radiation by people are increasing rapidly. If people could know how much they will be exposed by radiation, it may help them avoiding it and understand what exactly radiation is. By doing this, we were helping to reduce the anxiety of radiation contamination. In this study, we have researched figures of radioactivity with 'Captus-3000 thyroid uptake measurement systems' in well counter detector system. The materials were measured with Briquette, Shiitake, Pollock, Button type battery, Alkaline battery, Topsoil, Asphalt, Gasoline, Milk powder, Pine, Basalt stone, Pencil lead, Wasabi, Coarse salt, Tuna(can) Cigar, Beer, and then we categorized those samples into Land resources, Water resources, Foodstuff and Etc (Beer classified as a water resources has been categorized into Foodstuff). Also, we selected the standard radiation source linear 137Cs to measure the sensitivity of well counter detector. After that, we took cpm(counter per minute) for the well counter detector of thyroid uptake system's sensitivity. Then we compared the results of each material's cpm and converted those results to Bq/kg unit. There were a little limitation with the measurement equipment because it has less sensitivity than other professional equipment like 'High purity germanium radiation detector'. Moreover, We didn't have many choices to decide the materials. As a result, there are macroscopic differences among the rates of material's spectrum. Therefore, it had meaningful results that showed how much each material had emitted radiation. To compare the material's cpm with BKG, we've compounded their spectrums. By doing that, we were able to detect some differences among the spectrums at specific peak section. Lastly, Button type battery, Alkaline Battery, Briquette, Asphalt and Topsoil showed high value. There were classified emitting high radiation Group A and emitted lower radiation Group B. The Group A, alkaline battery showed higher rate of radiation by 7.67 %, and Button type battery was yield 4.65 % higher rate than BKG. Additionally, Asphalt (8.03 %), Topsoil (3.76 %), Briquette (7.46 %) were yield for higher values. Several samples of the daily supplies were yield little higher, but it seems safe to use in daily lives. In the case of the 'Foodstuff', all of the samples were safe and they were under the radiation limits of the Ministry of Food and Drug Safety for Food; thus, we highly recommend this study to you as a reference of common daily routine.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.276-284
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• 1993

The ABB-CE System-80 reactor power cutback system(RPCS) is designed to enable continuous operation of the reactor without trip in the events of the loss of one of the two main feedwater pumps and loss of load, and thus improves plant availability in a cost effective manner. In this study expansion of RPCS has been investigated for continuous reactor operation without trip in the event of an inward control element assembly(CEA) deviation including a single rod drop. Under the expanded function of RPCS the control system will provide a rapid core power reduction on demand by releasing CEAs to drop into the core and reduce the turbine power, if necessary, to follow the reactor power variation. This design feature which is included as the new design features to be incorporated in the ABB-CE System-80+ meets the EPRI advanced light water reactor(ALWR) requirements. For this study core analysis models of System-80+ have been developed to simulate the nuclear steam supply system(NSSS) response as well as the RPCS initiation of rapid CEA insertion. The results of this study demonstrate that the reactor trip can be avoided in the event of inward CEA deviation including a single rod drop by the RPCS initiation and thus the plant availability and capacity factor would be increased.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.6
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• pp.655-662
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• 2014

This study conducted tensile tests on an Alloy 690TT tube at room temperature (RT) and at $343^{\circ}C$ using tube- and ring-type specimens to investigate the stress-strain behavior and tensile properties of a steam generator (SG) tube in the axial and circumferential directions at RT and at the design temperature of a nuclear power plant (NPP). The results of the axial tensile test showed that yield point phenomena appeared at both RT and $343^{\circ}C$, and serrated flow in the stress-strain curve appeared at $343^{\circ}C$. Yield and tensile strengths for both directions were clearly lower at $343^{\circ}C$ compared to RT; however, the elongations were approximately the same at both test temperatures. Regardless of the test temperature, the strengths in the circumferential direction were lower by approximately 5~10 % than those in the axial direction. In addition, the test data revealed that the reduction in the yield and tensile strengths of the Alloy 690TT SG tube with the test temperature was more significant than that estimated by the temperature correction factor of ASME Sec.II.

• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.137-149
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• 1996

The present study is to analyze an integral test, BETHSY test 6.9c, which represent loss of RURS accident under mid-loop operation. Both the pressurizer manway and the steam generator outlet plenum manway are opened as vent paths in order to prevent the system from pressurization by removing the steam generated in the core. The main purposes are to gain insights into the physical phenomena and identify sensitive parameters. Assessment of capability of CATHARE2 prediction can be established the effective recovery procedures using the code in an actual plant. Most of important physical phenomena in the experiment could be predicted by the CATHARE2 code. The peak pressure in the upper plenum is predicted higher than experimental value by 7 kPa since the differential pressure between the pressurizer and the surge line is overestimated. The timing of core uncovery is delayed by 500 seconds mainly due to discrepancy in the core void distribution. It is demonstrated that openings of the pressurizer manwey and the steam generator manway can prevent the core uncovery using only gravity feed injection. Although some disagreements are found in the detailed phenomena, the code prediction is considered reasonable for the overall system behaviors.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.3
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• pp.189-203
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• 2008

The physical and chemical properties of radioactive waste drums, which have been temporarily stored on site, should be characterized before their shipment to a disposal facility in order to prove that the properties meet the acceptance guideline. The investigation of NDT(Nondestructive Test) method was figured out that the contents in drum, the quantitative analysis of free standing water and void fraction can be examined with X-ray NDT techniques. This paper describes the characteristics of X-ray NDT such as its principles, the considerations for selection of X-ray system, etc. And then, the waste drum characteristics such as drum type and dimension, contents in drum, etc. were examined, which are necessary to estimate the optimal X-ray energy for NDT of a drum. The estimation results were that: $(R)\acute{A}$ the proper X-ray energy is under 3 MeV to test the drums of 320 ${\beta}\S$ and less; $(R)\ddot{E}$ both X-ray systems of 450 keV and/or 3 MeV might be needed considering the economical efficiency and the realization. The number of drums that can be tested with 450 keV and 3 MeV X-ray system was figured out as 42,327 and 18,105 drums (based on storage of 2006. 12), respectively. Four testing scenarios were derived considering equipment procurement method, outsourcing or not, etc. The economical and feasibility assessment for the scenarios was resulted in that an optimal scenario is dependent on the acceptance guide line, the waste generator's policy on the waste treatment and the delivery to a disposal facility, etc. For example, it might be desirable that a waste generator purchases two 450 keV mobile system to examine the drums containing low density waste, and that outsourcing examination for the high density drums, if all NDT items such as quantitative analysis for 'free standing water' and 'void fraction', and confirmation of contents in drum have to be characterized. However, one 450 keV mobile system seems to be required to test only the contents in 13,000 drums per year.