• Nuclear Engineering and Technology
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• v.36 no.4
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• pp.346-356
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• 2004

14 Pressurized Water Reactors (PWR) in Korea use a remote monitoring system (RMS), which have been used in Korea since 1998. A Memorandum of Understanding on Remote Monitoring, based on Enhanced Cooperation on PWRs, was signed at the 10th Safeguards Review Meeting in October 2001 between the International Atomic Energy Agency (IAEA) and Ministry Of Science and Technology (MOST). Thereafter, all PWR power plants applied for remote monitoring systems. However, the existing method is high cost (involving expensive telephone costs). So, it was eventually applied to an Internet system for Remote Monitoring. According to the Internet-based Virtual Private Network (VPN) applied to Remote Monitoring, the Korea Atomic Energy Research Institute (KAERI) came to an agreement with the IAEA, using a Member State Support Program (MSSP). Phase I is a Lab test. Phase II is to apply it to a target power plant. Phase III is to apply it to all the power plants. This paper reports on the penetration testing of Phase I. Phase I involved both domestic testing and international testing. The target of the testing consisted of a Surveillance Digital Integrated System (SDIS) Server, IAEA Server and TCNC (Technology Center for Nuclear Control) Server. In each system, Virtual Private Network (VPN) system hardware was installed. The penetration of the three systems and the three VPNs was tested. The domestic test involved two hacking scenarios: hacking from the outside and hacking from the inside. The international test involved one scenario from the outside. The results of tests demonstrated that the VPN hardware provided a good defense against hacking. We verified that there was no invasion of the system (SDIS Server and VPN; TCNC Server and VPN; and IAEA Server and VPN) via penetration testing.

• Nuclear Engineering and Technology
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• v.43 no.4
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• pp.317-328
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• 2011

Pyroprocessing technology was developed in the beginning for metal fuel treatment in the US in the 1960s. The conventional aqueous process, such as PUREX, is not appropriate for treating metal fuel. Pyroprocessing technology has advantages over the aqueous process: less proliferation risk, treatment of spent fuel with relatively high heat and radioactivity, compact equipment, etc. The addition of an oxide reduction process to the pyroprocessing metal fuel treatment enables handling of oxide spent fuel, which draws a potential option for the management of spent fuel from the PWR. In this context, KAERI has been developing pyroprocessing technology to handle the oxide spent fuel since the 1990s. This paper describes the current status of pyroprocessing technology development at KAERI from the head-end process to the waste treatment. A unit process with various scales has been tested to produce the design data associated with the scale up. A performance test of unit processes integration will be conducted at the PRIDE facility, which will be constructed by early 2012. The PRIDE facility incorporates the unit processes all together in a cell with an Ar environment. The purpose of PRIDE is to test the processes for unit process performance, operability by remote equipment, the integrity of the unit processes, process monitoring, Ar environment system operation, and safeguards related activities. The test of PRIDE will be promising for further pyroprocessing technology development.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.12
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• pp.1968-1973
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• 2004

Structural degradations are often experienced on the components of nuclear power plants in reactor pressure vessels (RPV) and steam generators (SG) when these components are exposed to high temperature and high pressure for a long period of time. Such conditions result in the change of microstructures and of mechanical properties of materials, which requires an evaluation of the safeguards related to structural integrity. In a primary reactor cooling system (RCS), a dissimilar weld zone exists between cast stainless steel (CF8M) in a pipe and low-alloy steel (SA508 cl.3) in a nozzle. Thermal aging is observed in CF8M as the RCS is exposed for a long period of time under the operating temperature between 290 and 33$0^{\circ}C$. Under the same conditions, it is well known that degradation is not observed in low alloy steel. An investigation of the effect of thermal aging on the various mechanical properties of the dissimilar weld zone is required. The purpose of the present investigation is to find the effect of thermal aging on the dissimilar weld zone. The specimens are prepared by an artificially accelerated aging technique maintained for various times at 43$0^{\circ}C$, respectively. Then, The various mechanical test for the dissimilar welds are performed.

• Mass Spectrometry Letters
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• v.2 no.4
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• pp.80-83
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• 2011

Isotope ratio analysis of nuclear materials in individual particles is of great importance for nuclear safeguards. Although secondary ion mass spectrometry (SIMS) and thermal ionization mass spectrometry (TIMS) are utilized for the analysis of individual uranium particles, few studies were conducted for the analysis of individual uranium-plutonium mixed oxide particles. In this study, we applied SIMS and inductively coupled plasma mass spectrometry (ICP-MS) to the isotope ratio analysis of individual U-Pu mixed oxide particles. In the analysis of individual U-Pu particles prepared from mixed solution of uranium and plutonium standard reference materials, accurate $^{235}U/^{238}U$, $^{240}Pu/^{239}Pu$ and $^{242}Pu/^{239}Pu$ isotope ratios were obtained with both methods. However, accurate analysis of $^{241}Pu/^{239}Pu$ isotope ratio was impossible, due to the interference of the $^{241}Am$ peak to the $^{241}Pu$ peak. In addition, it was indicated that the interference of the $^{238}UH$ peak to the $^{239}Pu$ peak has a possibility to prevent accurate analysis of plutonium isotope ratios. These problems would be avoided by a combination of ICP-MS and chemical separation of uranium, plutonium and americium in individual U-Pu particles.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.488-493
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• 1996

보장조치 대상 원자력 시선에 대한 사찰 목적은 평화적 목적으로 사용되기 위한 시설 및 핵물질이 핵무기 생산 등의 비평화적 목적으로 전용되지 않았음을 확인하는 것이다. 이를 위하여 국제원자력기구에서는 보장조치 기준(IAEA Safeguards Criteria : 1991 - 1995)에 따라 적절한 검증 수단을 사용하여 핵물질의 형태 및 양, 시설의 운전기록 등에 대하여 보고된 내용과 실제 상황과의 일치성을 확인하고, 미신고된 핵활동이 없음을 확인하고 있다. 보장조치 측면에서 보면, 중수형원자로(CANDU)는 핵연료의 크기가 작고 운전중에 핵연료를 교체하는 방식(On Load Reactors)을 채택하고 있기 때문에 시설 내에서의 핵물질 이동이 매우 빈번하며, 사용후핵연료의 양 역시 경수형원자로에 비해 매우 많다. 따라서 중수형원자로에 대한 보장조치 사찰은 경수형원자로에 비해 사찰일수(최대허용사찰량 : 중수형원자로 45 인-일/년, 경수형원자로 15 인-일/년)가 훨씬 많고 보장조치 관련 장비 또한 매우 다양하다. 현재 운전 중인 월성 1호기에 이어 건설 중인 월성 2, 3, 4호기의 운전이 시작되면 중수형원자로에 대한 국제원자력기구 및 국가사찰 양이 급격히 늘어날 전망이다. 또한 월성 1호기의 경우 사용후핵연료 저장조의 용량 초과로 인한 건식저장고(Dry Canister)로의 이송이 1992년도부터 매년 실시되고 있으며, 이 기간 중에 이송 대상 핵연료의 검증 및 운반 중 전용을 방지하기 위한 추가적인 사찰이 수행됨으로써 많은 인력과 시간이 투입되고 있다. 또한 국제원자력기구에서 추진하고 있는 보장조치 강화 방안의 일환으로 현재 건설 중인 월성 2, 3, 4호기에 대해서는 월성 1호기에는 적용되지 않은 추가적인 보장조치 관련 장비의 설치가 고려되고 있다. 이에 따라 우리나라에서는 중수형원자로에 대한 국제 원자력기구의 사찰 기준 및 사찰 내용을 분석, 중수형원자로 보장조치 사찰에 대한 개선점을 도출하고, 후속기에 대해서 보다 효율적이고 효과적인 보장조치 방안을 적용토록 하여야 할 것이다.

• Journal of Radiation Protection and Research
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• v.43 no.3
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• pp.120-123
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• 2018

Background: Analysis of enrichment of $UO_2$ is important to verify the information declared by the license-holders. The redundancy methods are required to guarantee the analysis result. Korea Institute of Nuclear Nonproliferation and Control (KINAC) used to analyze it with alpha spectrometry and consign to Korea Basic Science Institute (KBSI) Thermal Ionization Mass Spectrometry (TIMS). This article evaluated the similarity of the results with two methods and derive correlation equation. It could be compared to the results measured by TIMS running by KBSI. Materials and Methods: There are not many certified materials for the uranium enrichment value. Therefore, 34 uranium pellets, which have the wide range of uranium enrichment from 0.21 to 4.69 wt%, were used for the experiments by the alpha spectrometry and the TIMS. Results and Discussion: The study shows there are the tendency of analyzed enrichment by each equipment. It shows uranium enrichment with alpha spectrometry evaluated 17% higher than that with TIMS on average. The regression equations were also derived in case the similarity between the two results with two methods is lower than predicted. Two experiments were designed to compare the effect of number of samples. The $R^2$ was 0.9977 with 34 pellets. It shows the equation is appropriate to predict the enrichment values by TIMS with that of alpha spectrometry. The $R^2$ was 0.9858 with four pellets for ten times. The $R^2$ decreased while the number of samples increased. The discrepancy between the lowest and highest enrichment seems to be one of the reason for it. Conclusion: KINAC expects the first equation with 34 samples is useful to predict the result with TIMS, the redundancy method, based on the alpha spectrometry. The extra samples are necessary to collect if the enrichment value analyzed by TIMS is lower than the value predicted with the equation. Further study would be followed related to the impact of the peak counts for each uranium isotopes, sample amount and number of experiments when TIMS established in KINAC by the end of 2018.

• Journal of the Korean Society for Nondestructive Testing
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• v.19 no.2
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• pp.93-99
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• 1999

SCAI(spent CANDU fuel bundle serial number identifier) was developed to read serial numbers of spent fuel bundles in the spent fuel storage. For the purpose of effectively identifying the serial number of fuel bundle. SCAI was composed of underwater camera & light part. guiding & supporting part and control & monitor part. So it is easy to assemble and disassemble, and operate. It was tested to read serial numbers of spent fuel bundles loaded in basket during the recent spent fuel transfer campaign at Wolsong Unit 1. And it was also applied to read serial numbers of spent fuel bundles discharging from the initial core at Wolsong Unit 3 by slight change of camera and light. Inspectors could easily operate SCAI after several practices in the storage pond, which was a user friendly. And SCAI provided clear and immediate picture for identification of serial numbers of spent fuel bundles. It was interally evaluated that SCAI greatly contributed to cut inspection efforts for national and international safeguards at Wolsong power plant.

• Journal of Radiation Protection and Research
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• v.39 no.1
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• pp.54-60
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• 2014

IAEA has employed various types of radiation detectors - HPGe, NaI, CZT - for accountancy of nuclear material. Among them, HPGe has been mainly used in verification activities required for high accuracy. Due to its essential cooling component(a liquid-nitrogen cooling or a mechanical cooling system), it is large and heavy and needs long cooling time before use. New hand-held portable HPGe has been developed to address such problems. This paper deals with results of performance evaluation test of the new hand-held portable HPGe prototype which was used during IAEA's inspection activities. Radioactive spectra obtained with the new portable HPGe showed different characteristics depending on types and enrichments of nuclear materials inspected. Also, Gamma-rays from daughter radioisotopes in the decay series of $^{235}U$ and $^{238}U$ and characteristic x-rays from uranium were able to be remarkably separated from other peaks in the spectra. A relative error of enrichment measured by the new portable HPGe was in the range of 9 to 27%. The enrichment measurement results didn't meet partially requirement of IAEA because of a small size of a radiation sensing material. This problem might be solved through a further study. This paper discusses how to determine enrichment of nuclear material as well as how to apply the new hand-held portable HPGe to safeguard inspection. There have been few papers to deal with IAEA inspection activity in Korea to verify accountancy of nuclear material in national nuclear facilities. This paper would contribute to analyzing results of safeguards inspection. Also, it is expected that things discussed about further improvement of a radiation detector would make contribution to development of a radiation detector in the related field.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.237-247
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• 1993

Activity ratio of two radioactive primary fission products which had sufficiently different half-lives was expressed as functions of cooling time and irradiation histories in which average burnup, irradiation time, cycle interval time and the dominant fissile material of the spent fuel were included. The gamma-ray spectra of 36 samples from 6 spent PWR fuel assemblies irradiated in Kori unit-1 reactor were obtained by a spectrometric system equipped with a high purity germanium gamma-ray detector. Activity ratio $^{l44}$Ce $^{l37}$Cs, analyzed from each spectrum, was used for the calculation of cooling time. The results show that the radioactive fission products $^{l44}$Ce and $^{l37}$Cs are considered as useful monitors for cooling time determination because the estimated cooling time by detection of activity ratio $^{l44}$Ce $^{l37}$Cs agreed well with the operator declared cooling time within relative difference of $\pm$5 % despite the low counting rate of the gamma-ray of $^{l44}$Ce (about 10$^{-3}$ count per second). For the samples with several different irradiation histories, the determined cooling time by modeled irradiation history showed good agreement with that by known irradiation history within time difference of $\pm$0.5 year. From this result, it would be expected to be possible to estimate reliably the cooling time of spent nuclear fuel without the exact information about irradiation history. The feasibility study on identification of and/or sorting out spent nuclear fuel by applying the technique for cooling time determination was also performed and the result shows that the detection of activity ratio $^{l44}$Ce $^{l37}$Cs by gamma-ray spectrometry would be usefully applicable to certify spent nuclear fuel for the purpose of safeguards and management in a facility in which the samples dismantled or cut from spent fuel assemblies are treated, such as the post irradiation examination facility.mination facility.