During PIV (Physical Inventory Verification), the IAEA has been inspecting the CANDU-Type spent fuels using an optical fiber-based scintillation detector. KINAC has developed a new verification instrument to deal with problems of the existing one such as low sensitivity, heavy and large dimension, and inconvenience-in-use. Our previous studies focused on how to develop the new instrument and had not included its performance tests. Field tests were carried out recently at Wolsung unit 4 to evaluate performance of the existing and new instruments. The objective of this paper is to discuss background noise produced in the optical fiber signal cable itself. The verification equipment for the CANDU-type Heavy Water Reactor spent fuels uses a scintillation detector to bond a scintillation material to the end of an optical signal cable. At this time, the radiation signal obtained by a data acquisition system is the signal generated from the scintillator (p-terphenyl organic scintillator) and the optical signal cable ; The signal produced in the optical cable itself is background noise to degrade the spent fuel verification equipment. To characterize the background radiation noise, the spent fuel bundles at Wolsung Unit 4 were measured using the optical fiber cable without the radiation scintillator. This signal is generated by reaction of the optical cable and the radiation emitted from the spent fuel. From experimental results, it was observed that the background noise signal of the optical cable increased as the optical cable went down in the downward direction, because the cable length irradiated by the radiation increased with the optical cable area in the spent fuel storage pool. Difference in the background noise signal was dependent on the location of the vertical direction and the signal of the new optical cable was up to about 5 times higher than that of the existing cable. While, the new cable has the cross-section area about 3.2 times larger than the old cable. Our past studies showed that total signal amplitude – sum of signals generated from the scintillator and optical fiber - of the new verification instrument was at least about 15 times greater than that of the existing one. Considering the total signal and background noise signal, from this measured results, it was confirmed that the scintillator characteristics – in particular, light output and decay time – has a dominant impact on the signal sensitivity of the newly developed instrument. More details will be discussed at the conference.

Regulatory agencies require burn-up verification to ensure that dry storage casks using burn-up credit are not loaded with fuel with a reactivity greater than the allowable standard. Accordingly, in preparation for dry storage of SF, the reliability of the burnup was verified and action plans for fuel with confirmed errors were reviewed. Reliability verification was performed by comparing the actual burnup calculated with combustion calculation code (TOTE, ISOTIN) used in NPP and the design burnup calculated with the nuclear design code (ANC). As a result of comparing the differences between actual burnup and design burnup for 7,414 assemblies of SF generated from CE-type NPPs, the average deviation was confirmed to be 0.79% and 220 MWD/MTU. In the CE-type NPPs, no fuel showing large deviations was identified, and it was confirmed that reliability was secured. As a result of comparing the differences in 11,082 assemblies of SF generated from WH-type NPPs, the differences were not large, averaging 1.16% or 422 MWD/MTU. However, fuels showing significant differences were identified, and cause analysis was performed for those fuels. The cause analysis used a method of comparing the burnup of symmetrically loaded fuels in the reactor. For fuels that were not symmetrically loaded, a method was used to compare them with fuels with similar combustion histories. As a result of the review, it was confirmed that the fuel was under- or over-burned compared to symmetrically loaded fuel. For fuels for which clear errors have been identified, we are considering replacing them with the design burnup, and for fuels whose causes cannot be confirmed, we are considering ways to maintain the actual burnup.

More than 20,000 bundles of spent nuclear fuel are stored in the spent nuclear fuel storage pool of domestic nuclear power plants, and the dry storage facility project in the nuclear power plant site is being promoted as the saturation of the wet storage pool is imminent. Since bending or twisting of spent nuclear fuel is an important item in order to load spent nuclear fuel into a dry storage cask, PSE (Pool Side Examination) was performed to verify this. This paper describes whether it can be safely loaded into a dry storage cask based on the measurement results of bending or twisting of spent nuclear fuel. The nuclear fuel assembly is designed to prevent excessive assembly bending and twisting because it can cause interference during dry storage and handling due to factors such as differences in depletion of nuclear fuel rods, irradiation growth, and coolant flow during reactor operation. The bending of the nuclear fuel assembly is measured by establishing a Plumb Line to photograph the nuclear fuel assembly based on it, and calculating a pixel that images the distance between the support grid and the Plumb Line. The twisting of the nuclear fuel assembly is measured by forming a virtual vertical plane with two Plumb Lines, and based on this, the twisting angle of the lower fixed compared to the upper fixed. As a result of the measurement, the bending of spent nuclear fuel was about 0.0-10.2 mm, much lower than the reactor loading criteria of 15.0 mm, and in the case of twisting, about 0.0~2.2° much lower than the reactor loading criteria of 5.0°. Therefore, it was confirmed that spent nuclear fuel at domestic nuclear power plants was not affected by bending and twisting when loading into dry storage cask.

South It is necessary to develop the future technologies to improve the sustainability and acceptability of nuclear power plants generation. Currently, our company is preparing to build the dry storage facility on-site in accordance with the basic plan for managing high-level radioactive waste announced by the government in 2021. However, studies on technologies for the volume reduction of spent nuclear fuel to increase the efficiency of on-site spent fuel dry storage facilities are very not enough. Accordingly, in this study, the storage efficiency and appropriateness for the SF volume reduction processing technologies such as SF oxide processing technology and consolidation technology are evaluated. Finally, the goal is to develop the optimized technologies to improve the storage efficiency of spent nuclear fuel. As a result in this study is followings. [Safety] After removing volatile fission products (Xe, Kr, I, etc.), Xe, Kr, etc. are removed during storage of the sintered structures. UO2 has a high melting point of approximately 1,000°C after cesium (Cs) has been removed, and heat can be removed by natural convection. [Economy]1999 DUPIC unit facility unit price reference, 2020 standard 328 $/kg estimated. A Comprehensive Approach Considering the Whole System is needed. Benefit from replacement and continuous operation of metal storage containers. Changes in economic efficiency obtained in conjunction with fluctuations in electricity prices and disposal. [Waste filter] A separated solidification facility high-level waste filter is required, and overseas outsourcing must be considered. [Waste cladding]. Cannot be accommodated in low-level disposal site. This reason is why the Ni nuclides occur to be in bulk. [Metal structural material] It is possible to reduce the initial volume by 7.6% or more when compressed or melted, but the technology needs to be advanced. [Oxide blocks] Larger size and density are expected to improve storage and disposal efficiency. [Facilities operation waste] Expected to be able to be disposed of at mid-to-low level decommissioning sites in Gyeongju city. [Solidified volatile nuclides and activated metals] Expected to improve storage efficiency when used volume is reduced and stored, such as outsourced reprocessing. [Oxide block] Radioactivity and decay heat are estimated to be reduced by half during oxide treatment. 75% reduction in volume and 40% reduction in storage area compared to used nuclear fuel before treatment. [Merits/Shortages] Improvement of storage and disposal efficiency empirical research such as large-capacity [real-scale] oxide block production is required. Oxide processing facilities are likely to be classified as post-use nuclear fuel processing facilities. It is determined that additional documents such as a Radiation Environmental Report (RER) must be submitted. Existence of possible external leaks of glass, highly mobile radionuclides from the point of view of nuclear criticality and heat removal. Acceptancy requirements of citizens in the process of creating additional sites for oxide treatment facilities. Considering social public opinion, it is necessary to secure the acceptability such as residents’ opinions convergence. Characteristics of high nuclear non-propagation compared to other processing technologies involving chemical processing. Also, Expectation of volume reduction effect for spent nuclear fuel itself. Volume reduction methods for solid waste and gaseous waste are required.

Zircaloy-4 is utillzed in nuclear fuel rod cladding due to its strength and corrosion resistance. However, it can undergo deformation over time, known as creep, which poses a safety risk in reactors. Furthermore, hydrogen absorption during reactor operation can alter its properties and affect creep rates. Previous research suggests a trend in which hydrogen concentration corelates unidirectionally with creep rates, either increasing or decreasing as the concentration rises. This trend can also be observed in EPRI’s creep model, EDF-CEA Model-3. However, recent literature has suggested that creep behavior may vary depending on the state of hydrogen presence. Therefore, it has become evident that creep behavior can be influenced not only by hydrogen concentration but also by the state of hydrogen presence, whether it is in a solid solution state or precipitated as hydrides. Our study aimed to compare creep behavior in specimens with hydrogen concentrations below and above solubility limits. We fabricated Zircaloy-4 plate specimens with varying hydrogen concentrations and conducted creep tests. The results revealed that specimens below the solubility limit exhibited decreasing creep rates as hydrogen concentration increased, while those above the limit displayed increasing creep rates. This investigation confirms that the state of hydrogen presence significantly impacts creep behavior within Zircaloy-4 cladding. As part of our additional research plans, we intend to conduct creep tests on the material based on its orientation, whether it is in the rolling direction (RD) or the transverse direction (TD). We also plan to perform creep tests on ring specimens. Additionally, for the ring specimens, we aim to evaluate how creep behavior differs between the cold-worked stress-relieved (CWSR) condition and the recrystallized annealed (RXA) condition achieved through high-temperature heat treatment.

Any type of nuclear arms control or disarmament agreement requires some form of verification measure. Existing nuclear arms control treaties drew upon previous agreements such as the INF treaty, START, and IAEA nuclear safeguards inspections. However, previous treaties focused on limiting specific types of nuclear weapons and their delivery vehicles or reducing the total number of nuclear weapons rather than eliminating the nuclear enterprise as a whole. A potential nuclear disarmament verification treaty or agreement will depend on the geopolitical environment of the time as well as the national policies and priorities of each signatory state. Although research on the gradual reduction and eventual elimination of nuclear weapons is still ongoing, several states have cooperated to conduct experiments, exercises, and simulations on the procedures and technologies required for nuclear disarmament verification. Three of these efforts are the LETTERPRESS simulation conducted by the Quadrilateral Nuclear Verification Partnership (QUAD), NuDiVe Exercise conducted by the International Partnership for Nuclear Disarmament Verification (IPNDV), and the Menzingen experiment organized by the UNIDIR in partnership with the Swiss Armed Forces, Spiez Laboratory, Princeton University’s Program on Science and Global Security, and the Open Nuclear Network. These contain aspects for the development of a potential nuclear disarmament verification. The LETTERPRESS exercise conducted in 2017 tested potential activities and equipment inspectors might utilize in a nuclear weapon facility. The IPNDV NuDiVe exercises conducted in 2021 and 2022 tested the activities and equipment required for the verified dismantlement of a warhead within a dismantlement facility. Finally, the Menzingen experiment conducted in 2023 tested the practical procedures for the verification of a nuclear weapon’s absence at a storage site. This paper will analyze the three cases to offer considerations on the procedures and technologies future nuclear disarmament verification might include.

Arms control treaties during the Cold War generally used national technical means (NTM) to verify treaty compliance. This was because signatory states refused to agree on on-site inspection (OSI) measures since it would require some level of intrusion. Efforts on nuclear arms control such as the Limited Test Ban Treaty (LTBT) or Strategic Arms Limitation Talks (SALT) initially included some form of OSI but could not continue due to refusal from signatory states. The Intermediate-Range Nuclear Force (INF) treaty concluded between the US and the Soviet Union in 1978 was significant since both states agreed on a highly intrusive verification measure. The Strategic Arms Reduction Treaty (START) and the new START also called for OSI measures similar to the INF. Alongside reducing a significant number of nuclear warheads and limiting specific types of nuclear warhead delivery vehicles, these treaties also provided basic models for conducting on-site inspection (OSI). OSI measures primarily rely on the political agreement between signatory states. However, the structure, types of inspections, number of inspections allowed, and technology/equipment used in each of the regimes also differ according to the objectives of each treaty. The INF treaty and START are salient cases as basic models for current nuclear disarmament verification research. Thus, this paper will conduct a case study on the procedures and mechanisms required for nuclear arms control verification in terms of OSI. Using the implications drawn from the INF treaty and START, this paper offers considerations for a potential nuclear disarmament verification.

Safeguards systems and measures are determined through diversion scenario analysis based on the facility design information submitted to the IAEA when a new nuclear facility is introduced. While the concept of safeguards-by-design (SBD), which considers the safeguards from the design phase for a facility operator to minimize unplanned changes or disruption to facility operations as well as for the IAEA to increase the efficiency and effectiveness in safeguards implementation, has been emphasized for more than a decade, there is no practical tool or guidance on how to apply it. In this study, we develop a diversion path analysis tool and introduce how to apply SBD using it. A diversion path analysis tool was developed based on the elements that constitute diversion and the algorithm generated based on the initial information of facility and nuclear material flow. The results of utilizing the analysis tool depending on a different level of facility information and the safeguards set-ups were compared through examples. Taking a typical light water reactor as an example, the test analyzed the automatic generation of dedicated routes, configuration of safeguards measures, and diversion path analysis. Through this, the application and limitations of the analysis tool are discussed, and ideas for utilization according to the SBD concept and necessary regulatory guidance are proposed. The results of this study are expected to be directly utilized to domestic nuclear control during the regulation process for a construction of new nuclear power systems, and furthermore, to enhance national credibility in the engagement with the IAEA for implementation of safeguards.

We produced an activated carbon using sodium-lignosulfonate, in which we investigated how the sodium salt in lignin served as the activating agent during heat treatment. Our process resulted in a product with a high specific surface area of 1324 m2/ g at 800 °C and microporous structure. During the activation process, we observed the consumption of carbon due to the dehydration reaction of NaOH and the reduction of Na2CO3 to metallic Na, which created pores through oxidation/ reduction reactions. The intercalation of metallic Na between the lattices at high temperatures formed additional pores and increased the specific surface area. Our proposed mechanism holds promise for enhancing the control of the microstructure and porosity of activated carbons through the thermal treatment of biomass.

This paper proposes the optimal molecular weight for a petroleum-based binder pitch to enhance the density and strength of the prepared graphite block. The effect of the molecular weight on the binder properties, which was quantified using solvent fractionation, was considered based on the evaluation of the coking and viscosity characteristics. The affinity of the pitch to coke influenced the carbonization yield of the block, and the proportion of closed pores was reduced via the use of a highaffinity binder pitch. In addition, the viscosity was found to influence the uniformity of the coke and pitch dispersions, and numerous open pores were formed in the graphite block under high-viscosity conditions. In terms of the molecular weight, a reduction in the content of the insoluble 1-methyl-2-pyrrolidone (NMP) fraction, which was the heaviest fraction present in the pitch, was found to reduce the affinity of the binder to coke while increasing its viscosity. Therefore, the density and strength of the prepared graphite block were reduced upon increasing the insoluble NMP content of the binder pitch. Consequently, it was necessary to control the content of this fraction within < 13.81 wt% to obtain high-density and high-strength graphite blocks.

증산은 적정 관수 관리에 중요한 역할을 하므로 수분 스트레스에 취약한 토마토와 같은 작물의 관개 수요에 대한 지식이 필요하다. 관수량을 결정하는 한 가지 방법은 증산량을 측정하는 것인데, 이는 환경이나 생육 수준의 영향을 받는다. 본 연구는 분단위 데이터를 통해 수학적 모델과 딥러닝 모델을 활용하여 토마토의 증발량을 추정하 고 적합한 모델을 찾는 것을 목표로 한다. 라이시미터 데이터는 1분 간격으로 배지무게 변화를 측정함으로써 증산 량을 직접 측정했다. 피어슨 상관관계는 관찰된 환경 변수가 작물 증산과 유의미한 상관관계가 있음을 보여주었다. 온실온도와 태양복사는 증산량과 양의 상관관계를 보인 반면, 상대습도는 음의 상관관계를 보였다. 다중 선형 회귀 (MLR), 다항 회귀 모델, 인공 신경망(ANN), Long short-term memory(LSTM), Gated Recurrent Unit(GRU) 모델을 구 축하고 정확도를 비교했다. 모든 모델은 테스트 데이터 세트에서 0.770－0.948 범위의 R2 값과 0.495mm/min－ 1.038mm/min의 RMSE로 증산을 잠재적으로 추정하였다. 딥러닝 모델은 수학적 모델보다 성능이 뛰어났다. GRU 는 0.948의 R2 및 0.495mm/min의 RMSE로 테스트 데이터에서 최고의 성능을 보여주었다. LSTM과 ANN은 R2 값이 각각 0.946과 0.944, RMSE가 각각 0.504m/min과 0.511로 그 뒤를 이었다. GRU 모델은 단기 예측에서 우수한 성능 을 보였고 LSTM은 장기 예측에서 우수한 성능을 보였지만 대규모 데이터 셋을 사용한 추가 검증이 필요하다. FAO56 Penman-Monteith(PM) 방정식과 비교하여 PM은 MLR 및 다항식 모델 2차 및 3차보다 RMSE가 0.598mm/min으로 낮지만 분단위 증산의 변동성을 포착하는 데 있어 모든 모델 중에서 가장 성능이 낮다. 따라서 본 연구 결과는 온실 내 토마토 증산을 단기적으로 추정하기 위해 GRU 및 LSTM 모델을 권장한다.

Pseudoscorpiones De Geer, 1778, an order of arachnids, exhibits a global distribution spanning all continents, with the notable exception of Antarctica. This taxon is characterized by a wide array of unique behaviors, including phoresy and matriphagy, setting it apart from other arachnid groups. Moreover, their morphological features distinguish them notably from other animal taxa. Surprisingly, in the context of Korean research, pseudoscorpions remain a relatively unexplored field. Up until the year 2023, only 8 families, 13 genera, and 26 species have been documented in Korea. This number stands in stark contrast to neighboring countries like Japan, boasting 13 families, 35 genera, and an impressive 100 species, and China, which has documented 12 families, 39 genera, and an extensive 168 species of pseudoscorpions.

2022년부터 2023년까지 제주도내 키위 시설재배지를 대상으로 계절 초기 볼록총채벌레 발생 경향을 확인하 기 위해서 토양 표면의 잡초, 토양 표면 상단으로부터 60cm, 키위나무를 유인한 덕 상단 15cm에서 10일 간격으로 발생 조사하였다. 하우스 내부에서 발생하는 잡초 10종을 채집하여 조사한 결과, 갈퀴덩굴, 광대나물, 개불알풀, 별꽃, 뽀리뱅이, 황새냉이 6종에서 볼록총채벌레가 지속적으로 관찰되었다. 계절 초기 월동 성충의 비산시기를 확인하기 위해 토양 표면 60cm 위쪽에 설치한 황색 끈끈이트랩에서 2월 하순~3월 중순부터 볼록총채벌레의 발생을 확인하였다. 시설하우스 내부(덕 상단 15cm)와 외부(측장 높이)에 설치한 황색 끈끈이트랩을 비교해보면 시설 내·외부의 볼록총채벌레의 밀도가 증가하고 감소하는 시기가 유사하였다. 종합적인 고찰을 통하여, 발생 양상을 고려한 적절한 방제전략 수립이 요구된다.