• Nuclear Engineering and Technology
• /
• v.50 no.2
• /
• pp.313-317
• /
• 2018

Licenses for the storage of spent nuclear fuel (SNF) and vitrified highly active waste in casks under dry conditions are limited to 40 years and have to be renewed for prolonged storage periods. If such a license renewal has to be expected since as in accordance with the new site selection procedure a final repository for spent fuel in Germany will not be available before the year 2050. For transport and possible unloading and loading in new casks for final storage, the integrity and the maintenance of the geometry of the cask's inventory is essential because the SNF rod cladding and the cladding of the vitrified highly active waste are stipulated as a barrier in the storage concept. For SNF, the cladding integrity is ensured currently by limiting the hoop stress and hoop strain as well as the maximum temperature to certain values for a 40-year storage period. For a prolonged storage period, other cladding degradation mechanisms such as inner and outer oxide layer formation, hydrogen pick up, irradiation damages in cladding material crystal structure, helium production from alpha decay, and long-term fission gas release may become leading effects driving degradation mechanisms that have to be discussed.

• Nuclear Engineering and Technology
• /
• v.53 no.9
• /
• pp.2816-2829
• /
• 2021

Decay heat residuals of spent nuclear fuel (SNF), i.e., the differences between calculations and measurements, were obtained previously for various spent fuel assemblies (SFA) using the Polaris module of the SCALE code system. In this paper, we compare decay heat residuals to their uncertainties, focusing on four PWRs and four BWRs. Uncertainties in nuclear data and model inputs are propagated stochastically through calculations using the SCALE/Sampler super-sequence. Total uncertainties could not explain the residuals of two SFAs measured at GE-Morris. The combined z-scores for all SFAs measured at the Clab facility could explain the resulting deviations. Nuclear-data-related uncertainties contribute more in the high burnup SFAs. Design and operational uncertainties tend to contribute more to the total uncertainties. Assembly burnup is a relevant variable as it correlates significantly with the SNF decay heat. Additionally, burnup uncertainty is a major contributor to decay heat uncertainty, and assumptions relating to these uncertainties are crucial. Propagation of nuclear data and design and operational uncertainties shows that the analyzed assemblies respond similarly with high correlation. The calculated decay heats are highly correlated in the PWRs and BWRs, whereas lower correlations were observed between decay heats of SFAs that differ in their burnups.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.19 no.4
• /
• pp.517-532
• /
• 2021

Safe geological disposal of spent nuclear fuel (SNF) requires knowledge of the deep hydrochemical characteristics of the repository site. Here, we conducted a set of deep hydrochemical investigations using a 750-m borehole drilled in a model granite system in Wonju, South Korea. A closed investigation system consisting of a double-packer, Waterra pump, flow cell, and water-quality measurement unit was used for in situ water quality measurements and subsequent groundwater sampling. We managed the drilling water labeled with a fluorescein dye using a recycling system that reuses the water discharged from the borehole. We selected the test depths based on the dye concentrations, outflow water quality parameters, borehole logging, and visual inspection of the rock cores. The groundwater pumped up to the surface flowed into the flow cell, where the in situ water quality parameters were measured, and it was then collected for further laboratory measurements. Atmospheric contact was minimized during the entire process. Before hydrochemical measurements and sample collection, pumping was performed to purge the remnant drilling water. This study on a model borehole can serve as a reference for the future development of deep hydrochemical investigation procedures and techniques for siting processes of SNF repositories.

• Nuclear Engineering and Technology
• /
• v.54 no.8
• /
• pp.2819-2827
• /
• 2022

The disposal area of a deep geological repository (DGR) for the disposal of spent nuclear fuels (SNFs) is estimated considering the spacing between deposition holes and between disposal tunnels, as determined by a thermal analysis using the decay heat of a reference SNF. Given the relatively large amount of decay heat of the reference SNF, the disposal area of the DGR is found to be overestimated. Therefore, we develop a computer program using MATLAB, termed ACom (Assembly Combination), to combine SNFs when stored in canisters such that the decay heat per canister is evenly distributed. The stability of ACom was checked and the overall distribution of the decay heat per canister was analyzed. Finally, ACom was applied to disposal scenarios suggested in the conceptual design of a DGR for SNFs, and it was confirmed that the decay heat per canister could be evenly distributed and that the maximum decay heat of the canister could be much lower than that of a canister estimated using a reference SNF. ACom can be used to improve the disposal efficiency by reducing the disposal area of a DGR for SNFs by ensuringg a relatively even distribution of decay heat per canister.

• Nuclear Engineering and Technology
• /
• v.54 no.11
• /
• pp.4030-4048
• /
• 2022

In this study, road transportation tests were conducted with surrogate fuel assemblies under normal conditions of transport to evaluate the vibration and shock load characteristics of spent nuclear fuel (SNF). The overall test data analysis was conducted based on the measured acceleration and strain data obtained from the speed bump, lane-change, deceleration, obstacle avoidance, and circular tests. Furthermore, representative shock response spectrums and power spectral densities of each test mode were acquired. Amplification or attenuation characteristics were investigated according to the load transfer path. The load attenuated significantly as it transferred from the trailer to the cask. By contrast, the load amplified as it transferred from the cask to the surrogate SNF assembly. The fuel loading location on the cask disk assembly did not exhibit a significant influence on the strain measured from the fuel rods. The principal strain was in the vertical direction, and relatively large strain values were obtained in spans with large spacing between spacer grids. The influence of the lateral location of fuel rods was also investigated. The fuel rods located at the side exhibited relatively large strain values than those located at the center. Based on the strain data obtained from the test results, a hypothetical road transportation scenario was established. A fatigue evaluation of the SNF rod was performed based on this scenario. The evaluation results indicate that no fatigue damage occurred on the fuel rods.

• Nuclear Engineering and Technology
• /
• v.55 no.12
• /
• pp.4695-4702
• /
• 2023

Nuclear power plants play a pivotal role in the global energy infrastructure, fulfilling a substantial share of the world's energy requirements in a sustainable way. The management of these facilities, especially the handling of spent nuclear fuel (SNF), necessitates meticulous inspections to guarantee operational safety and efficiency. However, the prevailing inspection methodologies lean heavily on human operators, which presents challenges due to the potential hazards of the SNF environment. This study introduces the design of a novel Pre-Batched Inspection Strategy (PBIS) that integrates robotic automation and image processing techniques to bolster the inspection process. This methodology deploys robotics to undertake tasks that could be perilous or time-intensive for humans, while image processing techniques are used for precise identification of SNF targets and regulating the robotic system. The implementation of PBIS holds considerable promise in minimizing inspection time and enhancing worker safety. This paper elaborates on the structure, capabilities, and application of PBIS, underlining its potential implications for the future of nuclear energy inspections.

• Nuclear Engineering and Technology
• /
• v.56 no.3
• /
• pp.1100-1107
• /
• 2024

The nuclear waste attributes of near-term deployable SMRs were assessed using established nuclear waste metrics, which are the DU mass, SNF mass, volume, activity, decay heat, radiotoxicity, and decommissioning LLW volumes. Metrics normalized per unit electricity generation were compared to a reference large PWR. Three SMRs, VOYGR, Natrium, and Xe-100, were selected because they represent a range of reactor and fuel technologies and are active designs deployable by the decade's end. The SMR nuclear waste attributes show both some similarities to the PWR and some significant differences caused by reactor-specific design features. The DU mass is equivalent to or slightly higher than the PWR. Back-end waste attributes for SNF disposition vary, but the differences have a limited impact on long-term repository isolation. SMR designs can vary significantly in SNF volume (and thus heat generation density). However, these differences are amenable to design optimization for handling, storage, transportation, and disposal technologies. Nuclear waste attributes from decommissioning vary depending on design and decommissioning technology choices. Given the analysis results in this study and assuming appropriate waste management system and operational optimization, there appear to be no major challenges to managing SMR nuclear wastes compared to the reference PWR.

• Journal of the Korea Concrete Institute
• /
• v.17 no.5 s.89
• /
• pp.811-818
• /
• 2005

It is well known that the fluidity and the fluidity loss of fresh concrete are affected by the kind of organic admixtures. Organic admixture can improve the properties of concrete. Sulfonated Naphthalene-Formaldehyde(SNF) Superplasticizer is used representatively, but has a problem in fluidity loss. In this study, we synthesized the Sulfonated Melamine-Formaldehyde(SMF) superplasticizer at the various synthetic conditions and compared the physical properties with SMF superplasticizer. SW superplasticizer is synthesized with four synthetic steps. Step 1 is hydroxymethylation, Step. 2 is Sulfonation, Step. 3 is Polymerization and Step. 4 is Stabilization. Synthesis of SMF superplasticizer depends on pH, temperature and reaction time. In this reaction, we changed the mole ratio of melamine to formaldehyde at 1:3, 1:4, and the amount of acid catalyst at Step. 3. After application of SMF superplasticizer and its mixture with SNF superplasticizer to cement pastes and mortars, we measured the physical properties of them at the different dosages(0.5, 1.0, 1.5 wt%) to cement. All samples including superplasticizer showed higher compressive strengths and slump, smaller pore size and porosity than CEM

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.1
• /
• pp.87-93
• /
• 2006

Inventories, and characteristics such as dimension, fuel rod array, weight, $^{235}U$ enrichment, and discharge burnup of spent nuclear fuel (SNF) generated from existing and planed nuclear power plants based on National 2nd Basic Plan for Electric Power Demand and Supply were investigated and projected to support geological disposal system design. The historical and projected inventory by the end 2057 is expected to be 20,500 and 14,800MTU for PWR and CANDU spent nuclear fuel, respectively. The quantity of SNF with initial $^{235}U$ enrichment of 4.5 wt.% and below was shown to be 96.5% in total. Average burnup of SNF revealed $\sim36$ GWD/MTU and $\sim40$ GWD/MTU for the period of 1994-1999 and 2000-2003, respectively. It is expected that the average burnup of SNF will be $\sim45$ GWD/MTU at the end of 2000's. From the comprehensive study, it was concluded that the imaginary SNF with $16\times16$ Korean Standard Fuel Assembly, cross section of $21.4cm\times21.4cm$, length of 453cm, mass of 672 kg, initial $^{235}U$ enrichment of 4.5 wt.%, discharge burnup of 55 GWD/MTU could cover almost all SNFs to be produced by 2057.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2009.06a
• /
• pp.35-36
• /
• 2009

지속가능 에너지원인 원자력은 또한 글로벌 에너지원의 특성을 갖추고 있다 핵연료는 원자로에 장전되는 신규 핵연료를 구성하고 있는 우라늄 채광 단계에서부터 연소 후에 발생하는 사용후 핵연료 (SNF 또는 SF) 관리 단계에 이르는 전과정에 걸쳐 핵연료 사이클로 파악되고 있다. 이러한 전과정 관점에서 볼 때, 핵연료 사이클에 관여하는 이해관계자는 자국뿐만 아니라 다국가 (multination)를 포함하고 있다. 특히, 후행 핵연료 사이클인 사용후 핵연료의 저장, 처리, 처분 단계에서는 다국가 시나리오를 배제하지 않는 사용후 핵연료 관리 전략의 도입이 고려될 수 있다. 여기서 다국가는 접경 국가, 인접 국가, 핵연료 공급 국가, 재처리 제공 국가, 재처리 위탁 국가, SNF 통과 허용 국가, SNF 저장 부지 제공 국가, SNF 향후 이용 국가 등이 될 수 있다 [김성호 2006]. 현재 우리나라에서는 여러 국가로부터 수입되고 있는 신규 핵연료 물질을 연소시켜 나온 사용후 핵연료를 부지내에 임시 저장하고 있다. 사용후 핵연료 발생량의 추산에 따르면, 2016년쯤에 현재 임시저장 용량이 포화될 예정이다 이러한 상황에서 다국가 시나리오를 포함한 관리 전략은 다국가 시나리오를 배제한 관리 전략과 다각적인 측면 에서 비교 검토될 필요성이 있다. 사용후 핵연료의 영구 처분장 부지 확보를 해결하기 위한 선행 단계로 공론화 단계가 지금 준비되고 있다. 예컨대, 단기 공론화 관리 방안의 하나로 비록 소극적인 입장이지만 타국 위탁 재처리 방식이 고려되고 있다 [KRS 2009] 이 연구에서는 단기적인 사용후 핵연료 관리 전략으로 여러 가지 다국가 수준의 저장, 처리, 처분 방식으로 바탕으로 다국가 시나리오들을 제안하려고 한다. 이들 다국가 시나리오를 포함한 관리 전략은 현재 다국가 시나리오를 배제한 국내 사용후 핵연료 처분장 부지 선정이 정치적/사회적 수용성 문제로 어려운 상황에 처할 경우에 해결책을 찾는 데에 기여하리라 본다. 또한, 부지 선정 단계에서 바라지 않는 난항이 나타나는 경우에 국가 차원의 한 대비책으로 다음을 제안한다: 한편으로는 자국 저장 시설이 추진되면서, 다른 편으로는 타국 저장 부지를 확보하는 전략이 검토되어야 한다. 이러한 이중 노선 (dual track) 전략은 여러 유럽 국가에서 이미 고려되고 있는 방안이다 [Greenpeace 2005] 다양한 다국성 정도 (a degree of multinationality) 의 저장, 처리, 처분 방식을 연결하는 가능한 다국가 시나리오 구조가 Fig.1에 제시되어 있다. 다국가 시나리오를 구성하는 기본 요소는 다음과 같다’ 1) 자국 임시 저장; 2) 자국 재처리; 3) 자국 중간 저장; 4) 자국 영구 처분; 5) 다국가 중간 저장; 6) 다국가 재처리; 7 ) 다국가 영구 처분. 이들 기본 요소들을 다국성 정도에 따라 결합하면 다양한 다국가 시나리오들이 얻어진다. 이들을 포함한 SNF 관리 전략은 크게는 1) 다국가 재처리 전략， 2) 다국가 저장 및 재처리 전략， 또는 3) 다국가 처분 전략 등으로 분류될 수 있다.