Dry Storage Of Spent Nuclear Fuel Research Articles

Degradation mechanisms in overpack concrete of spent nuclear fuel dry storage systems: A review

Dry storage systems are successfully used for interim storage of commercial spent nuclear fuel (SNF). Many of these systems consist of sealed, welded canisters placed inside concrete storage modules. Associated with the need for extending the operational periods of some of these systems, there are concerns that their concrete overpacks could degrade over time. This could lead to a loss of strength and durability. Potential degradation mechanisms include chemical attacks, concrete cancer, or freeze–thaw (F-T) cycles, making them prone to failure during extended periods of dry storage, especially when stressed during a hazardous event, such as an earthquake. The concrete mixtures that have been used in dry storage systems have not changed significantly since their introduction, and available mixtures often have the same ingredients used in conventional concrete. With the latest evolutions in material science, new concrete composites with ultra-high strength and outstanding thermal and mechanical properties can be utilized in such structures. Within a collaboration between the University of Idaho and the Idaho National Laboratory, enabled by funding provided by the Center of Advanced Energy Systems (CAES), a thorough review of SNF dry storage system overpack degradation observations were conducted. Further, the latest technologies in concrete composite materials with extraordinary strength, outstanding durability were identified. This paper presents and summarizes the most relevant mechanism of concrete deterioration in SNF systems, and identifies new materials to be used in future concrete mixtures to increase the service life of SNF dry storage systems and to lengthen inspection intervals.

An Analysis of Potentially Implementing Standardized Canisters in the U.S. Commercial Waste Management System

This analytical study focuses on loading standardized transportation, aging, and disposal canisters (STADs) at commercial reactor sites and subsequent transportation, e.g., to a consolidated interim storage facility (CISF). Specifically, the amount of spent nuclear fuel (SNF) available to load into STADs with varying deployment dates is explored, and the scenarios are compared with a scenario in which STADs are never loaded at reactor sites. Two key findings are that about half of the U.S. inventory of commercial SNF could be captured in STADs if they were fully deployed by 2035 and that the percentage of SNF available to load into STADs decreases as STAD deployment is delayed. In additional scenarios, the effects of shipping STADs directly from at-reactor spent fuel pools (SFPs) to a CISF are analyzed for a STAD full deployment year of 2035. A key finding from the analysis is that the dry storage of SNF in STADs at reactor sites can be minimized by direct shipment to a CISF from reactor site SFPs. However, minimizing dry storage at reactor sites means maximizing the receipt rate for STADs at a CISF, and there is likely a more optimal point between the two scenarios for an overall cost-effective operation of waste management systems.

THERMAL ANALYSIS OF DRY STORAGE AND TRANSPORTATION CASKS CASTOR USING COBRA-SFS

State Office for Nuclear Safety is a regulator body in the Czech Republic handling also licensing of storage and transportation casks (SCs) of the spent nuclear fuel (SNF). One of the main concerns for the dry storage of SNF is the safe removal of the residual heat resulting from the decay of fission products, actinides, and activated construction materials. There are many storage systems in use around the world with different storage configuration, fuel designs and boundary conditions that need to be evaluated. Two thermal models of SNF storage casks CASTOR 440/84 and CASTOR 440/84M were developed using the COBRA-SFS code, which is an internationally accepted and validated code used for licensing of dry storage casks and safety evaluation. The casks are in operation at the Nuclear Power Plant Dukovany for dry storage of the WWER-440 SNF. Both models were validated and used for evaluation of several problems in different configuration and particularly for licensing purposes at the SUJB.

Characterization of Physical and Chemical Conditions for Atmospheric Corrosion of Spent Nuclear Fuel Dry Storage Canisters

For long-term interim storage, spent nuclear fuel (SNF) is currently stored in stainless steel canisters at reactor sites. At near-marine sites, deposition and deliquescence of chloride-rich sea-salt aerosols may lead to localized canister corrosion and potentially, to penetration via chloride-induced stress corrosion cracking (CISCC). To assess the potential for CISCC, and to aid in development of appropriate corrosion testing protocols, expected physical and chemical conditions have been characterized for canisters at four near-marine independent spent fuel storage installations (ISFSIs) (Diablo Canyon, Maine Yankee, Calvert Cliffs, and Hope Creek).First, appropriate diurnal temperature and relative humidity (RH) cycles for corrosion testing were determined. Diurnal weather fluctuations result in variations in brine volume and solute concentration that may affect initiation, rate, or extent of canister surface corrosion and CISCC. To determine appropriate testing conditions, daily cycles in canister surface environment were estimated by taking local weather data and applying an equation of state for water [1] to calculate RH values at elevated canister surface temperatures. Results indicated that even slightly elevated temperatures (5-10°C) greatly reduced the diurnal range of RH experienced on the canister surface, and that previous experimental studies had used simulated diurnal RH ranges that were much too large.Second, dust samples collected from in-service canister surfaces were used to evaluate the composition of soluble salts present. While sea-salt aerosols were dominant at one ISFSI, at other sites, the soluble salts contained elevated concentrations of nitrate, sulfate, and calcium relative to sea-salts. Despite storage at near-marine locations, measured nitrate:chloride ratios in the dusts were as high as 1:1. Deliquescent brine compositions, calculated using the geochemical solubility and speciation code EQ3/6 [2], differed greatly from those produced by sea-salt aerosols. The importance of other anions (i.e. NO3 - and SO4 2 -) as possible corrosion inhibitors will be evaluated experimentally, using brine formulas based on the measured salt compositions.Third, scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM-EDS) analysis of the dust samples indicated that salt aerosols comprised only a small fraction of the dust, with the largest fraction consisting of detrital silicate mineral grains. Such grains would largely be inert but may affect corrosion processes by spreading deliquesced brine over the surface through capillary effects. Larger particles may also act a crevice-formers, providing occluded regions for corrosion initiation. Inert dust particle size distributions have been determined from the SEM images and are being used to develop experimental testing protocols to evaluate these effects. Median particle sizes were in the 4-7 µm range, with the largest particles generally less than 100 µm. For laboratory testing, silicate particles with diameters of 5 µm and 74 µm (200 mesh) will be used, as well as a mixture to assess combined effects.Diurnal variations in temperature and RH, and the presence of other anionic species, may significantly affect the corrosiveness of near-marine locations by impacting brine corrosivity (composition and concentration) and rates of corrosion initiation. Inert dust particles may promote corrosion initiation by creating crevices on the metal surface and will certainly affect the brine distribution on the surface and the resulting cathode size and cathode kinetics. These data provide the basis for developing realistic test matrices to evaluate these effects. Acknowledgements: SNL is managed and operated by NTESS under DOE NNSA contract DE-NA0003525. This document is SAND2020-5449 A.

Safety Assessment of Aircraft Crash Accident Into Spent Nuclear Fuel Dry Storage Facility – A Review With Focus on Structural Evaluation

Since the 1970s, aircraft crash accidents have been considered as one of the severest external events that should be evaluated for license application of nuclear reactors. After the 9.11 terrorist attacks, many countries have performed safety assessment against intentional or targeted aircraft crashes into nuclear related facilities. In some countries, assessment against targeted aircraft crash was enforced by regulation and considered an important task for license approval. Safety assessment against aircraft crash is a technically difficult task and many countries manage R&D programs to improve its reliability. In this paper, regulations of many countries regarding safety assessment against aircraft crash are summarized, separating regulations for accident aircraft crash and those for targeted aircraft crash. Research performed in various countries on safety assessment of nuclear facility against aircraft crash are summarized, with a focus on spent nuclear fuel dry storage facilities.

Probabilistic Assessment of Peak Cladding Hoop Stress and Hydrogen Content of PWR SNF Rod Cladding

This study presents two statistical models that were developed to estimate the expected peak cladding hoop stress (CHS) and the amount of hydrogen in pressurized water reactor (PWR) spent nuclear fuel (SNF) rod cladding. Peak CHS is caused by high rod internal pressure during vacuum drying performed when transferring SNF to dry storage. During in-reactor operation of PWR fuel, the rod cladding tends to corrode and uptake hydrogen. The hydrogen content and CHS control hydride-related cladding embrittlement at low material temperatures. The two methodologies developed in this study were used to create a generic rod database with information on PWR SNF conditions. This database provides information on 100 000 randomly selected rods that form part of the current U.S. SNF inventory. According to the statistical results, the expected hydrogen content of PWR rod cladding is in a sensitive interval that may facilitate hydride reorientation. However, the modeling results suggest that the expected peak CHS of the selected rods is significantly below 90 MPa, which is the estimated lower bound stress necessary to trigger significant radial hydride embrittlement in cladding after being cooled to room temperature. Further, the results indicate that hydride embrittlement due to excessive hydrogen in cladding is unlikely. Therefore, a low probability of hydride-related embrittlement of PWR SNF cladding currently stored in the U.S. inventory is anticipated, even under consideration of low cladding temperatures after long-term SNF dry storage.

Preventing Stress Corrosion Cracking of Spent Nuclear Fuel Dry Storage Canisters

Multi-purpose canisters (MPCs) are being employed for dry canister storage of spent fuel at nuclear plant sites as a temporary approach until an interim or permanent dry storage site(s) is available. For storage in coastal or lakeside regions and even humid environments, the corrosive nature of the moist air with entrained chlorides can make the welded regions of the canisters susceptible to pitting and chloride induced stress corrosion cracking (CISCC). In this report we evaluate CISCC lifetimes of welded 316L stainless steel canister plates showing in excess of 19 times increase of laser peened panel sections vs. those left as-welded using ASTM G36 (2013) accelerated corrosive testing. Specifically cracks never developed or propagated into the laser peening area. We also provide measurements of residual stress in test plates and related calculations of stress intensity and depth expected in the full canister geometry. We discuss the relevance of stress depth to pitting depth and crack growth rates. For this project welded 316L stainless steel panels were configured to MPC geometry and laser peened in the same manner as deployed for actual canisters. Our two-dimensional stress mapping shows that high energy laser peening provides deep (>5 mm) plastic compression. Literature information is not clear as to specific dependence of crack initiation and growth rates on variables including temperature, specifics of chlorine exposure, pit nucleation and residual stress intensity and depth. In our CISCC testing, although extensive cracking developed quickly (less than 18 hours) in the unpeened area, no cracking developed at all in the laser peened sections after 340 hour exposure. We further show that cracking that initiated and grew in unpeened regions arrested upon propagating to the laser peened boundary. We performed detailed finite element stress analysis (FEA) to evaluate the difference in expected retention of stress generated by the peening of test panels and the more constrained geometry of actual storage canisters. Our analysis shows that the 4 mm depth of stress measured in unconstrained test panels correlates to actual stress depth in the confined canister geometry of 6 mm, that is approximately 2 mm deeper thus satisfying a key stress depth requirement. The accelerated CISCC tests indicate that MPC design life can be dramatically improved by laser peening thereby making the MPCs a viable technical and economic solution to prevent CISCC. Following qualification testing and system configuration, a high energy laser peening process was deployed to peen welds of the dry fuel canisters for the San Onofre Nuclear Power Plant. The laser peening thereby helps ensure that the storage canisters will remain free from chlorine induced stress corrosion cracking.

Preliminary Design of the Forced Gas Drying System for Spent Nuclear Fuel Dry Storage

원자력발전소내 습식저장중인 사용후핵연료의 건식저장을 위해서는 캐니스터 내부에 사용후핵연료를 옮겨 담은 이후, 건 식저장 캐니스터 내장품이나 사용후핵연료 다발의 부식방지 및 피복관의 열화방지를 위해 모든 수분은 제거해야 한다. 캐 니스터 내부의 수분을 제거하는데 사용할 수 있는 기체강제순환 건조시스템 개발을 위한 연구개발이 진행중이다. 본 연구 에서는 기체강제순환 건조시스템 설계, 제작을 위한 예비설계를 수행하였다. 예비설계에는 캐니스터 내부 잔존수분 제거를 위한 건조사례조사, 건조관련 규격이나 표준, 건조합격기준, 건조장치구성, 현장요구분석, 습식저장중인 사용후핵연료 특성 을 포함하였다. 예비설계를 통하여 기체강제순환 건조시스템의 설계 개념도와 P&ID를 도출하였고, 이를 활용하여 건조시 스템 제작을 위한 상세설계를 수행할 것이다.

Review of Spent Nuclear Fuel Dry Storage Demonstration Programs in US

Review of Spent Nuclear Fuel Dry Storage Demonstration Programs in US

Development of the Vacuum Drying Process for the PWR Spent Nuclear Fuel Dry Storage

본 논문은 국내 원전의 습식저장조에 저장 중인 경수로형 사용후핵연료를 금속겸용용기를 이용해 건식으로 운영하기 위한운영공정을 개발하는 것이다. 국내 경수로형 원전의 사용후핵연료는 1990년대 초부터 습식으로 소내에서 운반을 한 경험은 많으나 건식으로 운전한 경험은 전혀 없는 실정이다. 이에 따라 금속겸용용기를 운영할 수 있는 세부 운영공정을 개발하였으며 주요 운영공정에서 금속겸용용기의 주요 구성품 및 사용후핵연료의 안전성이 유지됨을 확인하였다. 단기운영공정은 총 21시간 내에 이루어지도록 절차를 수립하였고 단계별로 허용운전 시간(15시간 습식공정, 3시간 배수공정, 그리고 3시간 진공공정)도 제시하였다.

Dry Storage Casks Monitoring by Means of Ultrasonic Tomography

Spent nuclear fuel (SNF) is one of the most hazardous types of nuclear power plant waste. This fact emphasizes the importance of careful handling and storage of SNF. There are two current state-of-the art technologies of SNF storage facility: wet and dry. It is important to mention that IAEA does not determine which kind of handling strategy should be chosen, however it is noted that dry storage of SNF could be used for one hundred years. Mining and Chemical Enterprise (MCE) is one of the leading Russian companies that deals exclusively with the dry storage of SNF. This company has implemented a long-term storage scheme. At the same time MCE faced the challenge of nondestructive monitoring of the degradation process of structural material of cask and its sealing with weld seam. Currently, X-ray testing is used for this purpose but in order to provide an effective nonradioactive method of monitoring MCE has initiated a collaborative R&D project with TPU supported by the Russian Government. Ultrasonic industrial tomography technique was proposed as the solution. The method is based on application of phased and sparse arrays transducer with real-time visualization algorithm. Received acoustic data is processed and realized by means of Sampling Phased Array technology which is a collaborative development of TPU and I-Deal Technology, GmbH. The multichannel ultrasonic set-up of immersion control was assembled for performing testing of seven experimental specimens with representative defects (side drill-holes, notches, natural welding flaws). X-ray tomography of high-resolution was chosen as the reference method. All indications were successfully reconstructed in B and C-scans and 3D image. The next step is to automate the monitoring procedure completely and to introduce an evaluation tool for current flaw state and prediction of its further behavior.

Effects of Source and Geometry Modeling on the Shielding Calculations for a Spent Nuclear Fuel Dry Storage Cask

Surface dose rate distribution over a spent nuclear fuel dry storage cask was realistically evaluated using the MONACO with Automated Variance Reduction using Importance Calculations (MAVRIC) computational sequence in the SCALE6 code system, with special emphasis on the effects of detailed modeling on the source term and cask geometry. The first storage cask in Taiwan has been fabricated and will be ready for loading of the designated spent fuels from Taiwan Power Company’s first nuclear power plant. A test run is scheduled for 2013.Neutron and gamma-ray source terms of the first batch of 56 spent fuels were determined one by one according to their specifications, burnup histories, and cooling times. The geometry of the cask was modeled in detail including the prescribed loading pattern of 56 spent fuels in the canister. MAVRIC was modified to allow specification of the source intensity and the axial distribution for each fuel bundle, and this resulted in a factor of 3 difference in the calculated surface dose rates from fuel gammas. The main purpose for such comprehensive and detailed modeling was to compare the results with a simplified model and to predict a dose rate distribution as realistically as possible in preparation for making a high-quality comparison with field measurements. In addition to checking assumptions adopted in the safety analysis report, the results of this study can provide useful guidance for the preparation of a health physics program during the test run and, more importantly, pave the way for establishing a valuable benchmark problem.