Modeling Heat Transfer in Spent Fuel Transfer Cask Neutron Shields: A Challenging Problem in Natural Convection

2010 ◽  
Author(s):  
James A. Fort ◽  
Judith M. Cuta ◽  
Chris S. Bajwa ◽  
Emilio Baglietto
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Steady State ◽  
Natural Convection ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
The United States ◽  
Challenging Problem ◽  
Spent Fuel ◽  
Neutron Shield

In the United States, commercial spent nuclear fuel is typically moved from spent fuel pools to outdoor dry storage pads within a transfer cask system that provides radiation shielding to protect personnel and the surrounding environment. The transfer casks are cylindrical steel enclosures with integral gamma and neutron radiation shields. Since the transfer cask system must be passively cooled, decay heat removal from the spent nuclear fuel canister is limited by the rate of heat transfer through the cask components, and natural convection from the transfer cask surface. The primary mode of heat transfer within the transfer cask system is conduction, but some cask designs incorporate a liquid neutron shield tank surrounding the transfer cask structural shell. In these systems, accurate prediction of natural convection within the neutron shield tank is an important part of assessing the overall thermal performance of the transfer cask system. The large-scale geometry of the neutron shield tank, which is typically an annulus approximately 2 meters in diameter but only 10–15 cm in thickness, and the relatively small scale velocities (typically less than 5 cm/s) represent a wide range of spatial and temporal scales that contribute to making this a challenging problem for computational fluid dynamics (CFD) modeling. Relevant experimental data at these scales are not available in the literature, but some recent modeling studies offer insights into numerical issues and solutions. However, the geometries in these studies, and for the experimental data in the literature at smaller scales, all have large annular gaps that are not prototypic of the transfer cask neutron shield. This paper proposes that there may be reliable CFD approaches to the transfer cask problem, specifically coupled steady-state solvers or unsteady simulations; however, both of these solutions take significant computational effort. Segregated (uncoupled) steady state solvers that were tested did not accurately capture the flow field and heat transfer distribution in this application. Mesh resolution, turbulence modeling, and the tradeoff between steady state and transient solutions are addressed. Because of the critical nature of this application, the need for new experiments at representative scales is clearly demonstrated.

scholarly journals The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

MRS Advances ◽  
2018 ◽  
Vol 3 (19) ◽  
pp. 991-1003 ◽  
Author(s):  
Evaristo J. Bonano ◽  
Elena A. Kalinina ◽  
Peter N. Swift
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
United States Of America ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
The United States ◽  
Spent Fuel ◽  
Dry Storage ◽  
The Us

ABSTRACTCurrent practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-century when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.

Developing a Concept for a National Used Fuel Interim Storage Facility in the United States

2013 ◽  
Author(s):  
Donald Wayne Lewis
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Spent Nuclear Fuel ◽  
The United States ◽  
Yucca Mountain ◽  
Storage Facility ◽  
Spent Fuel ◽  
Geological Repository ◽  
Interim Storage

In the United States (U.S.) the nuclear waste issue has plagued the nuclear industry for decades. Originally, spent fuel was to be reprocessed but with the threat of nuclear proliferation, spent fuel reprocessing has been eliminated, at least for now. In 1983, the Nuclear Waste Policy Act of 1982 [1] was established, authorizing development of one or more spent fuel and high-level nuclear waste geological repositories and a consolidated national storage facility, called a “Monitored Retrievable Storage” facility, that could store the spent nuclear fuel until it could be placed into the geological repository. Plans were under way to build a geological repository, Yucca Mountain, but with the decision by President Obama to terminate the development of Yucca Mountain, a consolidated national storage facility that can store spent fuel for an interim period until a new repository is established has become very important. Since reactor sites have not been able to wait for the government to come up with a storage or disposal location, spent fuel remains in wet or dry storage at each nuclear plant. The purpose of this paper is to present a concept developed to address the DOE’s goals stated above. This concept was developed over the past few months by collaboration between the DOE and industry experts that have experience in designing spent nuclear fuel facilities. The paper examines the current spent fuel storage conditions at shutdown reactor sites, operating reactor sites, and the type of storage systems (transportable versus non-transportable, welded or bolted). The concept lays out the basis for a pilot storage facility to house spent fuel from shutdown reactor sites and then how the pilot facility can be enlarged to a larger full scale consolidated interim storage facility.

scholarly journals Design Of Dry Cask Storage For Serpong Multipurpose Reactor Spent Nuclear Fuel

2018 ◽  
Vol 24 (1) ◽  
Author(s):  
Dyah Sulistyani Rahayu ◽  
Yuli Purwanto ◽  
Zainus Salimin
Keyword(s):  
Heat Transfer ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Construction Material ◽  
Trial And Error ◽  
Spent Fuel ◽  
Concrete Wall ◽  
Transfer Resistance ◽  
Steel Liner ◽  
Fuel Elements

DESIGN OF DRY CASK STORAGE FOR SERPONG MULTI PURPOSE REACTOR SPENT NUCLEAR FUEL. The spent nuclear fuel (SNF) from Serpong Multipurpose Reactor, after 100 days storing in the reactor pond, is transferred to water pool interim storage for spent fuel (ISFSF). At present there are a remaining of 245 elements of SNF on the ISSF,198 element of which have been re-exported to the USA. The dry-cask storage allows the SNF, which has already been cooled in the ISSF, to lower its radiation exposure and heat decayat a very low level. Design of the dry cask storage for SNF has been done. Dual purpose of unventilated vertical dry cask was selected among other choices of metal cask, horizontal concrete modules, and modular vaults by taking into account of technical and economical advantages. The designed structure of cask consists of SNF rack canister, inner steel liner, concrete shielding of cask, and outer steel liner. To avoid bimetallic corrosion, the construction material for canister and inner steel liner follows the same material construction of fuel cladding, i.e. the alloy of AlMg2. The construction material of outer steel liner is copper to facilitate the heat transfer from the cask to the atmosphere. The total decay heat is transferred from SNF elements bundle to the atmosphere by a serial of heat transfer resistance for canister wall, inner steel liner, concrete shielding, and outer steel liner respectedly. The rack canister optimum capacity of 34 fuel elements was designed by geometric similarity method basedon SNF position arrangement of 7 x 6 triangular pitch array of fuel elements for prohibiting criticality by spontaneous neutron. The SNF elements are stored vertically on the rack canister.  The thickness of concrete wall shielding was calculated by trial and error to give air temperature of 30 oC and radiation dose on the wall surface of outer liner of 200 mrem/h. The SNF elements bundles originate from the existing racks of wet storage, i.e. rack canister no 3, 8 and 10. The value of I0 from the rack no 3, 8 and 10 are 434.307; 446.344; and 442.375 mrem/h respectively. The total heat decay from rack canister no 3,8 and 10 are 179.640 ; 335.2; and 298.551 watts. The result of the trial and error calculation indicates that the rack canister no 3, 8 and 10 need the thickness of concrete shielding of 0.1912, 0.1954 and 0.1940 m respectively.Keywords: heat and radiation decay, spent fuel , storage cask.

Thermal and Fluid Analysis of Test Canister for Spent Nuclear Fuel

2020 ◽  
Author(s):  
Marwan Charrouf ◽  
Allen Williams
Keyword(s):  
Nuclear Fuel ◽  
Heat Dissipation ◽  
Spent Nuclear Fuel ◽  
Flow Behavior ◽  
The United States ◽  
Time Profile ◽  
Electrical Heating ◽  
Cfd Modeling ◽  
Electric Heater ◽  
Spent Fuel

Abstract The absence of a long-term solution for the storage of spent nuclear fuel prompts utilities in the United States to establish on-site storage for used fuel. The challenges associated with placement of spent fuel in dry cask storage on the power plant’s Independent Spent Fuel Storage Installations (ISFSI’s) include aging management of the stainless steel canisters and monitoring for the possible onset of stress corrosion cracking (SCC). The San Onofre Nuclear Generating Station (SONGS) has initiated a test program to examine the effects of heat generation variations inside a test canister using an electric heater rather than spent fuel on the shell temperatures. The test helps in the evaluation of external environmental factors and shell temperature, and to monitor for SCC. This paper presents the computational fluid dynamics (CFD) modeling developed in support of the test to analyze the air natural circulation in the subgrade enclosure and within the test canister with the electrical heating. The thermal analysis is performed using ANSYS CFX and integrally simulates the flow behavior and heat transfer mechanisms both inside and outside the test canister. Comparison of results from different heat loads that represent the decay heat time-profile, sensitivity to the turbulence model, and modes of heat dissipation are discussed. The CFD results are also compared to in-situ temperature measurements to validate the analysis.

A Comparative Evaluation of Licensing Requirements for Dry Storage of Spent Nuclear Fuel in the United States, Germany, Canada and the Russian Federation

2004 ◽  
Author(s):  
Edward Wonder ◽  
David S. Duncan ◽  
Eric A. Howden
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
The United States ◽  
Approval Process ◽  
Spent Fuel ◽  
Regulatory Requirements ◽  
Fuel Storage ◽  
The Russian Federation ◽  
The Republic

Technical activities to support licensing of dry spent nuclear fuel storage facilities are complex, with policy and regulatory requirements often being influenced by politics. Moreover, the process is often convoluted, with numerous and diverse stakeholders making the licensing activity a difficult exercise in consensus-reaching. The objective of this evaluation is to present alternatives to assist the Republic of Kazakhstan (RK) in developing a licensing approach for a planned Dry Spent Fuel Storage Facility. Because the RK lacks experience in licensing a facility of this type, there is considerable interest in knowing more about the approval process in other countries so that an effective, non-redundant method of licensing can be established. This evaluation is limited to a comparison of approaches from the United States, Germany, Russia, and Canada. For each country considered, the following areas were addressed: siting; fuel handling and cask loading; dry fuel storage; and transportation of spent fuel. The regulatory requirements for each phase of the process are presented, and a licensing approach that would best serve the RK is recommended.

Reprocessing of Spent Nuclear Fuel: A Policy Analysis

2006 ◽  
Author(s):  
Todd P. Lagus
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Energy Content ◽  
The United States ◽  
Spent Fuel ◽  
Cycle System ◽  
Spent Fuel Reprocessing ◽  
The Cost ◽  
Fuel Reprocessing

Nuclear spent fuel reprocessing has lately reemerged as a subject of debate in the energy policy world. Since a 1977 Presidential Directive which deferred reprocessing of spent nuclear fuel (SNF), the United States has utilized a “once through” or “open cycle” system of nuclear fuel processing, which leaves most of the energy content in uranium unused. Current reprocessing technology increases the cost of nuclear electricity while only offering limited storage benefits. Advanced technologies have the potential to increase proliferation resistance, the need for more geologic repositories, and allow the United States to regain an international presence in reprocessing technology. The United States should not immediately engage in spent fuel reprocessing, but should begin aggressive research and development for new reprocessing technologies.

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