• 한국안전학회지
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• 제26권4호
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• pp.87-95
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• 2011

The urgent VAI(Vital Area Identification) method development is required since 'The Act of Physical Protection and Radiological Emergency' that is established in 2003 requires an evaluation of physical threats in nuclear facilities and an establishment of physical protection in Korea. The KAERI(Korea Atomic Energy Research Institute) has developed the VAI methodology and VAI software called as VIPEX(Vital area Identification Package EXpert) for identifying the vital areas. This study is to demonstrate the applicability of KAERI's VAI methodology to a hypothetical facility, and to identify the importance of information of cable and piping runs when identifying the vital areas. It is necessarily needed to consider cable and piping runs to determine the accurate and realistic TEPS(Top Event Prevention Set). If the information of cable and piping runs of a nuclear power plant is not considered when determining the TEPSs, it is absolutely impossible to acquire the complete TEPSs, and the results could be distorted by missing it. The VIPEX and FTREX(Fault Tree Reliability Evaluation eXpert) properly calculate MCSs and TEPSs using the fault tree model, and provide the most cost-effective method to save the VAI and physical protection costs.

• Nuclear Engineering and Technology
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• 제33권2호
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• pp.156-165
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• 2001

A radiological safety assessment was performed for a hypothetical near-surface radioactive waste repository as a simple screening calculation to identify important nuclides and to provide insights on the data needs for a successful demonstration of compliance. Individual effective doses were calculated for a conservative ground water pathway scenario considering well drilling near the site boundary. Sensitivity of resulting ingestion dose to input parameter values was also analyzed using Monte Carlo sampling. Considering peak dose rate and assessment time scale, C-14 and T-129 were identified as important nuclides and U-235 and U-238 as potentially important nuclides. For C-14, the dose was most sensitive to Darcy velocity in aquifer The distribution coefficient showed high degree of sensitivity for I-129 release.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.1108-1115
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• 2024

Assuring physical security for Micro Modular Reactors (MMRs) will be key to their licensing. Economic constraints however require changes in how the security objectives are achieved for MMRs. A promising new approach is the so-called performance based (PB) approach wherein the regulator formally sets general security objectives and leaves it to the licensee to set their own specific acceptance criteria to meet those objectives. To implement the PB approach for MMRs, one performs a consequence-based analysis (CBA) whose objective is to study hypothetical malicious attacks on the facility, assuming that intruders take control of the facility and perform any technically possible action within a limited time before an offsite security force can respond. The scenario leading to the most severe radiological consequences is selected and studied to estimate the limiting impact on public health. The CBA estimates the total amount of radionuclides that would be released to the atmosphere in this hypothetical scenario to determine the total radiation dose to which the public would be exposed. The predicted radiation exposure dose is then compared to the regulatory dose limit for the site. This paper describes application of the CBA to four different MMRs technologies.

• 방사성폐기물학회지
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• 제10권3호
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• pp.143-149
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• 2012

한국원자력연구원의 지하처분연구시설인 KURT 부지에 가상의 심지층 처분 시설을 가정하고 안전성평가를 수행하기 위해 필요한 지하수 유동 자료를 작성하기 위한 지하수 유동 모의가 수행되었다. 연구지역의 전반적인 지하수 유동 특성을 고려하기 위해, 광역 규모의 지하수 유동 모의를 먼저 실시하여 국지 규모 지하수 유동 모의에서 이용될 경계 조건을 구하고, 현장에서 확인된 단열 자료를 반영하여 국지 규모에서의 지하수 유동계가 모의되었다. 같은 방식으로 국지 규모에서 지하수 유동에 관한 경계 조건을 뽑아내어 KURT 부지 규모의 지하수 유동 모의에 이용하였다. 국지 규모의 지하수 유동 모의 결과로 얻어진 지하수위 분포를 통해 입자 추적(particle tracking) 모의를 수행하여 가상의 처분 부지 위치에서 지표로 흐르는 지하수의 유동 경로를 확인하고, 경로의 길이와 지하수의 시간당 유동량(discharge rate)을 구하였다. 본 연구에서 이용된 일련의 지하수 유동 모의 및 입자 추적 모의 방법은 향후 심지층 처분 시설의 안전성 평가에 필요한 자료를 작성하는데 유용하게 쓰일 것으로 기대된다.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3604-3613
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• 2022

The Physical Protection System (PPS) plays an important role and must effectively deal with various adversary attacks in nuclear security. In specific single adversary path scenarios, we can calculate the PPS effectiveness by EASI (Estimated Adversary Sequence Interruption) through Probability of Interruption (P_I) calculation. EASI uses a single value of the probability of detection (P_D) and the probability of alarm communications (P_C) in the PPS. In this study, we develop a multi-path analysis code based on EASI to evaluate the effectiveness of PPS. Our quantification method for P_I considers the variability and uncertainty of P_D and P_C value by Monte Carlo simulation. We converted the 2-D scheme of the nuclear facility into an Adversary Sequence Diagram (ASD). We used ASD to find the adversary path with the lowest probability of interruption as the most vulnerable paths (MVP). We examined a hypothetical facility (Hypothetical National Nuclear Research Facility - HNNRF) to confirm our code compared with EASI. The results show that implementing the variability extension can estimate the P_I value and its associated uncertainty. The multi-path analysis code allows the analyst to make it easier to assess PPS with more extensive facilities with more complex adversary paths. However, the variability of the P_D value in each protection element allows a significant decrease in the PI value. The possibility of this decrease needs to be an important concern for PPS designers to determine the P_D value correctly or set a higher standard for PPS performance that remains reliable.

• 한국국방경영분석학회지
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• 제9권2호
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• pp.23-30
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• 1983

A simulation study was conducted to determine optimum capacity of Navy drydock facility using GASP-IV, an advanced FORTRAN-based simulation language, under demands of regular overhauls and emergency repairs by ships of an hypothetical fleet composition. Three year dock usage data was analyzed to produce probability distributions underlying drydock repair demands. The present facility size of two drydocks was simulated and was found to be somewhat short of adequate service capability, showing excessive average waiting time and average queue length. The simulation model was then modified to include an additional drydock of similar size as the other two and a year's simulation was again conducted. All repair needs were quite satisfactorily met and all docks showed very high utilization factor (0.98). This contributed to an increase in the fleet's ship availability from 0.95 to 0.99. This study illustrates the usefulness of simulation technique as a tool for analyzing policy alternatives in military long-term investment areas.

• Journal of Radiation Protection and Research
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• 제34권1호
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• pp.1-7
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• 2009

It is expected that a substantial amount of spent fuels will be transported from the four nuclear power plant (NPP) sites in Korea to a hypothetical centralized interim storage facility or a final repository in the near future. The cost for the transportation is proportional to the amount of spent fuels. In this paper, a cost estimation program is developed based on the conceptual design of a transportation system and a logistics analysis. Using the developed computer program, named as CASK, the minimum capacity of a centralized interim storage facility (CISF) and the transportation cost for PWR spent fuels are calculated. The PWR spent fuels are transported from 4 NPP sites to a final repository (FR) via the CISF. Since NPP sites and the CISF are located along the coast, a sea-transportation is considered and a road-transportation is considered between the CISF and the FR. The result shows that the minimum capacity of the interim storage facility is 15,000 MTU.

• 한국가스학회지
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• 제20권3호
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• pp.59-65
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• 2016

가스 사용 시설에서 폭발 위험성 평가 등급에 따라 적합한 방폭용 설비를 사용하는 것은 매우 중요하다. 가스 관련 법에서 가스 사용시설의 방폭 기준은 제시하고 있으나 폭발 위험장소 구분을 위한 기술 기준은 별도로 제시되어 있지 않다. 본 연구에서 한국산업표준 KS를 이용하여 저압 도시가스 배관시설에 대해 합리적인 폭발위험성 예측 방법을 제시하고자 한다. 누출공 크기, 누출압력에 따른 가상체적, 환기 유효성 등의 중요변수를 적용하여 폭발위험성이 예측되었다. 자연 환기 조건을 만족하는 실험 설비가 제작되어 도시 가스 누출 실험 결과와 KS 표준에 의해 예측된 폭발 위험성 예측 결과가 비교되었다.

• 한국안전학회지
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• 제29권2호
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• pp.24-30
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• 2014

There have been controversies over whether explosion hazardous area(EHA) should be classified for facilities which use lighter-than-air gases such as city gas, hydrogen and ammonia. Two view points are confronting each other: an economic piont of view that these gases are lighter than air and disperse rapidly, hence do not form EHA upon release into the atmosphere, and a safety point of view that they are also inflammable gases, hence can form EHA although the extent is limited compared to heavy gases. But various standards such as KS, IEC, API, NFPA do not exclude light gases when classifying EHA and present examples of EHA for light gas facilities. This study calculates EHA using the hypothetical volume in the IEC code where the hole sizes required for the calculation were selected according to various nominal pipe sizes in such a way to conform to the EHA data in the API code and HSL. Then, 25 leakage scenarios were suggested for 5 different pipe sizes and 5 operating pressures that cover typical operating conditions of light gas facilities. The EHA for the minimum leakage scenario(25 mm pipe, 0.01MPa pressure) was found to correspond to a hypothetical volume larger than 0.1 $m^3$(medium-level ventilation). This confirms the validity of classifying EHA for facilities using lighter-than-air gases. Finally, a computer program called HACPL was developed for easy use by light gas facilities that classifies EHA according to operating pressures and pipe sizes.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.380-387
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• 2015

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as $^{235}U$, $^{239}Pu$, $^{241}Pu$, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux ($>10^{12}n/cm^2{\cdot}s$) neutron source comprised of a high-energy (30 MeV)/high-current (~2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (< $0.06{\mu}Sv/h$), a few shielding materials [high-density polyethylene (HDPE)eBorax, $B_4C$, and $Li_2CO_3$] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.