Shipping Cask Research Articles

Monte Carlo Shielding Calculations for a Spent Fuel Transport Cask with Automated Monte Carlo Variance Reduction

For the purpose of performing reasonable shielding calculation of a spent fuel transport cask, the use of Monte Carlo methods has been proposed for solving the radiation transport problem on a detailed structure of the transport cask considering fixed neutron sources. A SMIRE (Simplified MCNP-ANISN_W Variance Reduction) system has been developed in the present study, which is possible to generate automatically the lower weight boundary of the weight window for each mesh based on the Consistent Adjoint Driven Importance Sampling (CADIS). Compared with the case of the importance based on the empirical formula, the figure of merit is increased by a factor of 25. In this system, it is possible to calculate the weight suitable for the distantly-positioned detector point from the fuel effective region to introduce the relaxation factor which relaxes the increase of the particle numbers at the boundary of weight window meshes that are generally caused by the large attenuation of adjoint flux. This system is used for a variety of the radiation transport problems as well as the transport cask.

Application of Dose Evaluation of the MCNP Code for the Spent Fuel Transport Cask

Application of Dose Evaluation of the MCNP Code for the Spent Fuel Transport Cask

사용후핵연료 운반용기 덮개 내부 열전달 해석

사용후핵연료 운반용기 표면온도가 <TEX>$85^{\circ}C$</TEX>를 초과할 경우, 대인용 보호막(Personnel Barrier) 또는 운반용 덮개(Transport Hood)를 설치하여 운반 중 운반용기 표면에 사람이 직접 접근할 수 없도록 하여야 한다. 운반용 덮개가 설치된 경우, 열적 안전성 평가의 한 가지 경우인 정상조건 열해석 시, 외부환경 경계조건(환경온도 및 외부복사온도)으로 적용하기 위해서 운반용 덮개 내부 열 환경 조건(내부 공기온도 및 운반용 덮개 표면온도)을 계산해야 한다. 따라서 본 연구에서는 운반용 덮개 내부 공기온도 및 표면온도를 계산하기 위한 해석적 방법 및 열전달 특성에 대한 분석을 수행하였고 CFD 해석 결과와 비교를 통해 타당성을 검증하였다. In case that the maximum temperature of any surface readily accessible during transport of a spent nuclear fuel (SNF) transport cask exceeds <TEX>$85^{\circ}C$</TEX> in the absence of insolation under the ambient temperature of <TEX>$38^{\circ}C$</TEX>, personnel barriers or transport hood shall be used to prevent people from casual contact with the transport cask surface. Usually the air temperature within the hood and the hood surface temperature are calculated and further utilized as boundary conditions(free stream temperature and external radiation temperature) for thermal evaluation under normal conditions of transport. In this study, these temperatures are derived using the analytical method based on the heat transfer mechanism around the transport cask under transport hood assuming the thermal equilibrium. By comparing the analytical solutions with the results from the detailed calculations with CFD-computer-code FLUENT 12.1 it is verified that the analytical method is still efficient tool to estimate the temperatures and these temperatures can be further used as boundary conditions for thermal evaluation under normal conditions of transport.

연소도이득효과(BUC) 적용 사용후핵연료 운반용기의 임계 불확실도 평가

In general, conventional criticality analyses for spent fuel transport/dry storage systems have been performed based on assumption of fresh fuel concerning the potential uncertainties from number density calculation of Transuranic and Fission Products in spent fuel. However, because of economic loss due to the excessive criticality margin, recently the design of transport/dry storage systems with Burnup Credit(BUC) application has been actively developed. The uncertainties in criticality analyses on transport/storage systems with BUC technique show strong dependance upon initial enrichment and burnup rate, whereas those in the conventional criticality evaluation based on fresh fuel assumption do not show such a dependance. In this study, regulatory-required uncertainties of the criticality analyses for BK 26 Cask, which is conceptually designed spent fuel transport cask with BUC corresponding to the limiting circumstances on nuclear power plants in Korea, are evaluated as a function of initial enrichment and burnup rate. Results of this study will be used as basic data for spent fuel loading curve of BK 26 Cask.

Response of nuclear materials in storage and transport to malevolent environments

AbstractWork related to the assessment of radiological health consequences resulting from a sabotage attack on nuclear fuel storage or transport casks has been continuing since the late 1970s. While the level of effort in this area has been uneven over these three decades due to policy priorities, funding levels and programmatic priorities of the countries funding this type of work, substantial progress has been made. From phenomenology of in cask transport processes to development of aerosol production in high energy attack environments, the analytical and experimental work performed provides substantial justification to consequence assessments that heretofore have had to rely on conservative assumptions in lieu of empirical data. One constant since the late 1990s in addressing this problem has been an international working group whose primary focus has been to develop source term data from experimental simulations of sabotage types of attacks. This working group, titled the International Working Group f...

A CFD analysis of thermal behaviour of transportation cask under fire test conditions

Thermal design of transportation cask for shipping radioactive waste needs strict compliance with the guidelines of the regulatory bodies. Lead shielding is usually provided in these casks for arresting gamma rays and reducing hazardous emissions to the environment below permissible limits. During transportation, accidental fire may break out and cause melting of lead for a prescribed duration. The present analysis reports, for the first time, a comprehensive CFD analysis of the thermal behaviour of melting of lead under high Rayleigh number convection during the fire test. The study reveals a substantial influence of natural convection on the thermal state and melting behaviour of lead which may have a great bearing on the safety and security of public during transportation of cask.

Integrity Evaluation and Dynamic Behavior of Radioactive Material Transport Cask under Drop Impact

Radioactive waste is shipped in casks designed to prevent the release of the radioactive material into the environment in normal accidental conditions. So, the cask must maintain its integrity such as shielding characteristics and leak tightness after the 9m free drop. In this paper, we consider several types of radioactive material transport casks. The drop scenarios for the integrity evaluation are 6 drop cases. The dynamic analysis is performed by applying the various free drop angle of cask for research the dynamic impact behavior and integrity evaluation of a radioactive material transport cask. As a results of impact drop analysis on radioactive material transport cask, the container maintain structural integrity evaluation in 9 meter free drop conditions.

Finite element mesh design of a cylindrical cask under puncture drop test conditions

Transport casks for radioactive materials have to withstand the 9 m drop test, 1 m puncture drop test and dynamic crush test with regard to the mechanical requirements according to the IAEA regulations. The safety assessment of the package can be carried out on the basis of experimental investigations with prototypes or models of appropriate scale, calculations, by reference to previous satisfactory safety demonstrations of a sufficiently similar nature or a combination of these methods. Computational methods are increasingly used for the assessment of mechanical test scenarios. However, it must be guaranteed that the calculation methods provide reliable results. Important quality assurance measures at the Federal Institute for Materials Research and Testing are given concerning the preparation, run and evaluation of a numerical analysis with reference to the appropriate guidelines. Hence, a successful application of the finite element (FE) method requires a suitable mesh. An analysis of the 1 m puncture ...

Experience with MIDUS, a type B(U)99molybdenum transport package

AbstractThe medical isotope 99molybdenum (99Mo) decays to 99mtechnetium (99mTc), and 99mTc is the most widely used radioactive isotope for nuclear medicine diagnostic studies for the heart, kidneys, lungs, liver, spleen and bone, among others. A major supplier of 99Mo replaced its fleet of 99Mo transport casks with a new type B(U) package designed and manufactured by EnergySolutions, and certified by the US Nuclear Regulatory Commission. The packages have been in service for 3 years, and an operational database is now available. This paper discusses the particular challenges for the Medical Isotope Depleted Uranium Shielded Transport Package (MIDUS) design and licensing, and how the field performance of MIDUS has compared with design objectives. The MIDUS package was designed to have approximately twice the capacity as its predecessor, but with the same outside envelope dimensions. Depleted uranium shielding was necessary in order to meet size and capacity requirements. Bulk shielding was not the only des...

Industrial research for transmutation scenarios

This article presents the results of research scenarios for americium transmutation in a 22nd century French nuclear fleet, using sodium fast breeder reactors. We benchmark the americium transmutation benefits and drawbacks with a reference case consisting of a hypothetical 60 GWe fleet of pure plutonium breeders. The fluxes in the various parts of the cycle (reactors, fabrication plants, reprocessing plants and underground disposals) are calculated using EDFʼs suite of codes, comparable in capabilities to those of other research facilities. We study underground thermal heat load reduction due to americium partitioning and repository area minimization. We endeavor to estimate the increased technical complexity of surface facilities to handle the americium fluxes in special fuel fabrication plants, americium fast burners, special reprocessing shops, handling equipments and transport casks between those facilities.

Measurement and Analysis of In-vessel Component Activation and Gamma Dose Rate Distribution in Joyo

th periodic inspection of the experimental fast reactor Joyo of the Japan Atomic Energy Agency, the irradiation test subassembly (S/A) was bent and became an obstruction in the reactor vessel (R/V), and the upper core structure (UCS) was damaged. This incident made it necessary to perform an in-vessel visual inspection using an optical fiber scope. This inspection revealed that the UCS needed to be replaced for Joyo re-start. A critical task in restoring Joyo was evaluating the activation of the UCS; this was necessary to design the transportation cask for the damaged UCS. However, it is difficult to directly obtain samples of the highly irradiated components in the R/V. In order to obtain experimental data and to evaluate the accuracy of radiation dose calculation, in-vessel gamma dose rate measurements were conducted in the experimental fast reactor Joyo. The measured gamma dose rate was compared with the values calculated using the QAD code and the ORIGEN2 code with the JENDL-3.2 cross section library.

Developments in Back End of the Fuel Cycle of Indian Thermal Reactors

Abstract Substantial progresses made in last five decades in the front end of the nuclear fuel cycle in India has been supported adequately with advances and innovations in the field of back end technology and methodology developments for power reactors in general and thermal reactors in particular. The activities in the back end are characterized by spent fuel transport and storage, reprocessing, waste management and large scale support technology developments. The specifically designed spent fuel transport casks are in operation for shipment of the spent fuel to pre-reprocessing underwater storage facilities operating at Tarapur and Kalpakkam. Three reprocessing plants based on PUREX process located at Trombay, Tarapur and Kalpakkam have been set up to undertake reprocessing of the spent fuel from Trombay based research reactors, power reactors of western sector and southern sector respectively. The complete waste treatment, immobilization and disposal cycle for Low and Intermediate Level waste is in place to manage such waste from power as well as research reactors with a number of chemical treatment plants and near surface disposal facilities operational in various power plant sites. As regards to management of high level liquid wastes mainly arising from the reprocessing stream, two vitrification plants are currently operational whereas third one in under commissioning. One pre-disposal storage facility for vitrified high level waste products is operational at Tarapur while another new storage facility is under construction at Kapakkam. Research and development activities in the field of permanent geological disposal of such vitrified high level waste are in progress. These activities in the back end have been amply supported by development of large technological innovations and design of remote handling systems, robotics, radiation shielding etc. Development of advanced matrices, innovations in processes related to actinide separations and partitioning, recovery of other valuable isotopes are among the areas of future research and developments in fuel cycle of thermal reactors in India.

Monte Carlo Shielding Comparative Analysis Applied to TRIGA HEU and LEU Spent Fuel Transport

The paper is a comparative study of LEU and HEU fuel utilization effects for the shielding analysis during spent fuel transport. A comparison against the measured data for HEU spent fuel, available from the last stage of spent fuel repatriation fulfilled in the summer of 2008, is also presented. All geometrical and material data for the shipping cask were considered according to NAC-LWT Cask approved model. The shielding analysis estimates radiation doses to shipping cask wall surface, and in air at 1 m and 2 m, respectively, from the cask, by means of 3D Monte Carlo MORSE-SGC code. Before loading into the shipping cask, TRIGA spent fuel source terms and spent fuel parameters have been obtained by means of ORIGEN-S code. Both codes are included in ORNL's SCALE 5 programs package. The actinides contribution to total fuel radioactivity is very low in HEU spent fuel case, becoming 10 times greater in LEU spent fuel case. Dose rates for both HEU and LEU fuel contents are below regulatory limits, LEU spent fuel photon dose rates being greater than HEU ones. Comparison between HEU spent fuel theoretical and measured dose rates in selected measuring points shows a good agreement, calculated values being greater than the measured ones both to cask wall surface (about 34% relative difference) and in air at 1 m distance from cask surface (about 15% relative difference).

ANALYSIS OF THE TRANSPORTATION LOGISTICS FOR SPENT NUCLEAR FUEL IN KOREA

As a part of the back-end fuel cycle, transportation of spent nuclear fuel (SNF) from nuclear power plants (NPPs) to a fuel storage facility is very important in establishing a nuclear fuel cycle. In Korea, the accumulated amount of SNF in the NPP pools is troublesome since the temporary storage facilities at these NPP pools are expected to be full of SNF within ten years. Therefore, Korea cannot help but plan for the construction of an interim storage facility to solve this problem in the near future. Especially, a decision on several factors, such as where the interim storage facility should be located, how many casks a transport ship can carry at a time and how many casks are initially required, affect the configuration of the transportation system. In order to analyze the various possible candidate scenarios, we assumed four cases for the interim storage facility location, three cases for the load capacity that a transport ship can carry and two cases for the total amount of casks used for transportation. First, this study considered the currently accumulated amount of SNF in Korea, and the amount of SNF generated from NPPs until all NPPs are shut down. Then, how much SNF per year must be transported from the NPPs to an interim storage facility was calculated during an assumed transportation period. Second, 24 candidate transportation scenarios were constructed by a combination of the decision factors. To construct viable yearly transportation schedules for the selected 24 scenarios, we created a spreadsheet program named TranScenario, which was developed by using MS EXCEL. TranScenario can help schedulers input shipping routes and allocate transportation casks. Also, TranScenario provides information on the cask distribution in the NPPs and in the interim storage facility automatically, by displaying it in real time according to the shipping routes, cask types and cask numbers that the user generates. Once a yearly transportation schedule is established, TranScenario provides some statistical information, such as the voyage time, the availability of the interim storage facility, the number of transported casks sent from the NPPs, and the number of transported casks received at the interim storage facility. By using this information, users can verify and validate a yearly transportation schedule. In this way, the 24 candidate scenarios could be constructed easily. Finally, these 24 scenarios were compared in terms of their operation cost.

Comparison of the transportation risks for the spent fuel in Korea for different transportation scenarios

According to the long term management strategy for spent fuels in Korea, they will be transported from the spent fuel pools in each nuclear power plant to the central interim storage facility (CISF) which is to start operation in 2016. At the start of the operation of the final repository (FR), by the year 2065, transport will then take place between the CISF and the FR. Therefore, we have to determine the safe and economical logistics for the transportation of these spent fuels by considering their transportation risks and costs. In this study, we developed four transportation scenarios for a maritime transportation by considering the type of transportation casks and transport means in order to suggest safe and economical transportation logistics for the spent fuels in Korea. And, we estimated and compared the transportation risks for these four transportation scenarios. Also, we estimated and compared the transportation risks resulting from accidents during the transportation of PWR and PHWR spent fuels by road trailers from the CISF and the FR. From the results of this study, we found that risks resulting from accidents during the transportation of the spent fuels have a very low radiological risk activity with a manageable safety and health consequences. The results of this study can be used as basic data for the development of safe and economical logistics for a transportation of the spent fuels in Korea by considering the transportation costs for the four scenarios which will be needed in the near future.

Thermal analysis of proposed transport cask for three advanced burner reactor used fuel assemblies

Preliminary studies of used fuel generated in the US Department of Energy's Advanced Fuel Cycle Initiative have indicated that current used fuel transport casks may be insufficient for the transportation of said fuel. This work considers transport of three 5-year-cooled oxide advanced burner reactor used fuel assemblies with a burn-up of 160 MWD kg–1. A transport cask designed to carry these assemblies is proposed. This design employs a 7-cm-thick lead gamma shield and a 20-cm-thick NS-4-FR composite neutron shield. The temperature profile within the cask, from its centre to its exterior surface, is determined by two-dimensional computational fluid dynamics simulations of conduction, convection and radiation within the cask. Simulations are performed for a cask with a smooth external surface and various neutron shield thicknesses. Separate simulations are performed for a cask with a corrugated external surface and a neutron shield thickness that satisfies shielding constraints. Resulting temperature profiles indicate that a three-assembly cask with a smooth external surface will meet fuel cladding temperature requirements but will cause outer surface temperatures to exceed the regulatory limit. A cask with a corrugated external surface will not exceed the limits for both the fuel cladding and outer surface temperatures.

Buckling of fuel rods under inertia loading

AbstractThe buckling analysis of fuel rods during an end drop impact of a spent fuel transportation cask has traditionally been performed to demonstrate the structural integrity of the fuel rod cladding or the integrity of the fuel geometry in criticality evaluations for a cask drop event. The actual calculation of the fuel rod buckling load, however, has been the subject of some controversy, with estimates of the critical buckling load differing by as much as a factor of 5. Typically, in the buckling analysis of a fuel rod, assumptions are made regarding the percentage of fuel mass that is bonded to or that participates with the cladding during the buckling process, with estimates ranging from 0 to 100%. The greater the percentage of fuel mass that is assumed to be bonded to the cladding, the higher the inertia loads on the cladding, and, therefore, the lower the 'g' value at which buckling occurs. However, these solutions do not consider displacement compatibility between the fuel and the cladding durin...

New approach to evaluate lattice expansion of light water reactor fuel elements on criticality safety of transport packages under impact accidents

AbstractIn order to safely transport packages containing light water reactor fuel assemblies, it is essential to maintain the fuel assemblies in a subcritical state in accidents during transport. To evaluate nuclear criticality safety, an estimator is required to determine an absolutely safe level based not only on hypothetical accidents but also on practical accident levels which, to some extent, are based on actual accidents. The purpose of the present study is to suggest the arrangement of the deformation range of the fuel assembly after an actual accident, and to obtain the maximum value of the neutron effective multiplication factor based on the criticality safety assessment for the transport cask. In the present study, two kinds of criticality calculations for the package were considered: large scale pin pitch shift and small scale pin pitch shift. For the large scale pin pitch shift, a parameter which determines the location of each fuel pin which constitutes the fuel assembly was introduced so tha...

Challenges in the renewal of the Swedish sea transport system

AbstractSince 1985, SKB has successfully operated a sea transport system for transport of spent nuclear fuel and radioactive waste to the intermediate storage facility, Clab and the final repository, SFR, in Sweden. The main components in the system are the ship M/S Sigyn, transport casks for spent fuel and core components, IP2 containers and terminal vehicles.

Spent fuel transport cask thermal evaluation under normal and accident conditions

The casks used for transport of nuclear materials, especially the spent fuel element (SPE), must be designed according to rigorous acceptance criteria and standards requirements, e.g. the International Atomic Energy Agency ones, in order to provide protection to people and environment against radiation exposure particularly in a severe accident scenario. The aim of this work was the evaluation of the integrity of a spent fuel cask under both normal and accident scenarios transport conditions, such as impact and rigorous fire events, in according to the IAEA accident test requirements. The thermal behaviour and the temperatures distribution of a Light Water Reactor (LWR) spent fuel transport cask are presented in this paper, especially with reference to the Italian cask designed by AGN, which was characterized by a cylindrical body, with water or air inside the internal cavity, and two lateral shock absorbers. Using the finite element code ANSYS a series of thermal analyses (steady-state and transient thermal analyses) were carried out in order to obtain the maximum fuel temperature and the temperatures field in the body of the cask, both in normal and in accidents scenario, considering all the heat transfer modes between the cask and the external environment (fire in the test or air in the normal conditions) as well as inside the cask itself. In order to follow the standards requirements, the thermal analyses in accidents scenarios were also performed adopting a deformed shape of the shock absorbers to simulate the mechanical effects of a previous IAEA 9 m drop test event. Impact tests on scale models of the shock absorbers have already been conducted in the past at the Department of Mechanical, Nuclear and Production Engineering, University of Pisa, in the ‘80s. The obtained results, used for possible new licensing approval purposes by the Italian competent Authority of the cask for PWR spent fuel cask transport by the Italian competent Authority, are discussed.