scholarly journals BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

2015 ◽  
Author(s):  
Bruce Balkcom Bevard ◽  
Ugur Mertyurek ◽  
Randy Belles ◽  
John M. Scaglione
Keyword(s):  
Research And Development ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Interim Storage

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 Recent Progress on the Standardized DOE Spent Nuclear Fuel Canister

2002 ◽  
Author(s):  
D. Keith Morton ◽  
Spencer D. Snow ◽  
Tom E. Rahl ◽  
Tom J. Hill ◽  
Richard P. Morissette
Keyword(s):  
Nuclear Fuel ◽  
Recent Progress ◽  
Spent Nuclear Fuel ◽  
Transportation Systems ◽  
Pressure Vessels ◽  
Department Of Energy ◽  
Regulatory Agencies ◽  
Spent Fuel ◽  
American Society ◽  
Interim Storage

The Department of Energy (DOE) has developed a set of containers for the handling, interim storage, transportation, and disposal in the national repository of DOE spent nuclear fuel (SNF). This container design, referred to as the standardized DOE SNF canister or standardized canister, was developed by the Department’s National Spent Nuclear Fuel Program (NSNFP) working in conjunction with the Office of Civilian Radioactive Waste Management (OCRWM) and the DOE spent fuel sites. This canister had to have a standardized design yet be capable of accepting virtually all of the DOE SNF, be placed in a variety of storage and transportation systems, and still be acceptable to the repository. Since specific design details regarding the storage, transportation, and repository disposal of DOE SNF were not finalized, the NSNFP recognized the necessity to specify a complete DOE SNF canister design. This allowed other evaluations of canister performance and design to proceed as well as providing standardized canister users adequate information to proceed with their work. This paper is an update of a paper [1] presented to the 1999 American Society of Mechanical Engineers (ASME) Pressure Vessels and Piping (PVP) Conference. It discusses recent progress achieved in various areas to enhance acceptance of this canister not only by the DOE complex but also fabricators and regulatory agencies.

Long Term Storage of Nuclear Spent Fuel as Key Role of Japan’s Nuclear Fuel Cycle Until 2100: Cost and Benefit

2010 ◽  
Author(s):  
Tadahiro Katsuta
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
Storage Site ◽  
Spent Fuel ◽  
Final Disposal ◽  
Enriched Uranium ◽  
Interim Storage ◽  
Reprocessing Plant

Political and technical advantages to introduce spent nuclear fuel interim storage into Japan’s nuclear fuel cycle are examined. Once Rokkasho reprocessing plant starts operation, 80,000 tHM of spent Low Enriched Uranium (LEU) fuel must be stored in an Away From Reactor (AFR) interim storage site until 2100. If a succeeding reprocessing plant starts operating, the spent LEU will reach its peak of 30,000 tHM before 2050, and then will decrease until the end of the second reprocessing plant operation. Throughput of the second reprocessing plant is assumed as twice of that of Rokassho reprocessing plant, indeed 1,600tHM/year. On the other hand, tripled number of final disposal sites for High Level Nuclear Waste (HLW) will be necessary with this condition. Besides, large amount of plutonium surplus will occur, even if First Breeder Reactors (FBR)s consume the plutonium. At maximum, plutonium surplus will reach almost 500 tons. These results indicate that current nuclear policy does not solve the spent fuel problems but rather complicates them. Thus, reprocessing policy could put off the problems in spent fuel interim storage capacity and other issues could appear such as difficulties in large amount of HLW final disposal management or separated plutonium management. If there is no reprocessing or MOX use, the amount of spent fuel will reach over 115,000 tones at the year of 2100. However, the spent fuel management could be simplified and also the cost and the security would be improved by using an interim storage primarily.

Measurement of Atmospheric Sea Salt Concentration in the Dry Storage Facility of the Spent Nuclear Fuel

2006 ◽  
Author(s):  
Masumi Wataru ◽  
Hisashi Kato ◽  
Satoshi Kudo ◽  
Naoko Oshima ◽  
Koji Wada ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Sea Salt ◽  
Storage Facility ◽  
Japan Society ◽  
Spent Fuel ◽  
Dry Storage ◽  
Storage Methods ◽  
Interim Storage

Spent nuclear fuel coming from a Japanese nuclear power plant is stored in the interim storage facility before reprocessing. There are two types of the storage methods which are wet and dry type. In Japan, it is anticipated that the dry storage facility will increase compared with the wet type facility. The dry interim storage facility using the metal cask has been operated in Japan. In another dry storage technology, there is a concrete overpack. Especially in USA, a lot of concrete overpacks are used for the dry interim storage. In Japan, for the concrete cask, the codes of the Japan Society of Mechanical Engineers and the governmental technical guidelines are prepared for the realization of the interim storage as well as the code for the metal cask. But the interim storage using the concrete overpack has not been in progress because the evaluation on the stress corrosion cracking (SCC) of the canister is not sufficient. Japanese interim storage facilities would be constructed near the seashore. The metal casks and concrete overpacks are stored in the storage building in Japan. On the other hand, in USA they are stored outside. It is necessary to remove the decay heat of the spent nuclear fuel in the cask from the storage building. Generally, the heat is removed by natural cooling in the dry storage facility. Air including the sea salt particles goes into the dry storage facility (Figure 1). Concerning the concrete overpack, air goes into the cask body and cools the canister. Air goes along the canister surface and is in contact with the surface directly. In this case, the sea salt in the air attaches to the surface and then there is the concern about the occurrence of the SCC. For the concrete overpack, the canister including the spent fuel is sealed by the welding. The loss of sealability caused by the SCC has to be avoided. To evaluate the SCC for the canister, it is necessary to make clear the amount of the sea salt particles coming into the storage building and the concentration on the canister. In present, the evaluation on that point is not sufficient. In this study, the concentration of the sea salt particles in the air and on the surface of the storage facility are measured inside and outside of the building. For the measurement, two sites of the dry storage facility using the metal cask are chosen. This data is applicable for the evaluation on the SCC of the canister to realize the interim storage using the concrete overpack.

Japanese Codes and Standards for Dual Purpose Metal Casks for Spent Nuclear Fuel

2014 ◽  
Author(s):  
Toshiari Saegusa ◽  
Makoto Hirose ◽  
Norikazu Irie ◽  
Masashi Shimizu
Keyword(s):  
Nuclear Fuel ◽  
Power Plants ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Spent Fuel ◽  
Dual Purpose ◽  
Service Conditions ◽  
Interim Storage ◽  
And Storage

The first Japanese spent fuel interim storage facility away from a reactor site is about to be commissioned in Mutsu City, Aomori Prefecture. In designing, licensing and construction of the Dual Purpose Casks (DPCs, for transport and storage) for this facility, codes and standards established by the Atomic Energy Society of Japan (AESJ) and by the Japan Society of Mechanical Engineers (JSME) have been applied. The AESJ established the first standard for DPCs as “Standard for Safety Design and Inspection of Metal Casks for Spent Fuel Interim Storage Facilities” in 2002 (later revised in 2010). The standard provides the design requirements to maintain the basic safety functions of DPCs, namely containment, heat removal, shielding, criticality prevention and the structural integrity of the cask itself and of the spent fuel cladding during transport and storage. Inspection methods and criteria to ensure maintenance of the basic safety functions and structural integrity over every stage of operations involving DPCs including pre-shipment after storage are prescribed as well. The structural integrity criteria for major DPC components refer to the rules provided by the JSME. JSME completed the structural design and construction code (the Code) for DPCs as “Rules on Transport/Storage Packagings for Spent Nuclear Fuel” in 2001 (later revised in 2007). Currently, the scope of the rules cover the Containment Vessel, Basket, Trunnions and Intermediate Shell as major components of DPCs. Rules for these components are based on those for components of nuclear power plants (NPP) with similar safety functions, but special considerations based on their shapes, loading types and required functions are added. The Code has differences from that for NPP components with considerations to DPC characteristics; - The primary stress and the secondary stress generated in Containment Vessels shall be evaluated under Service Conditions A to D (from ASME Sec III, Div.1). - Stress generated in the seal region lid bolts of Containment Vessels shall not exceed yield strength under Service Conditions A to D in order to maintain the containment function. - Fatigue analysis on Baskets is not required, and Trunnions can be designed only for Service Conditions A and B with special stress limits consistent with conventional assessment methods for transport packages. - Stress limits for earthquakes during storage are specified. - Ductile cast iron with special fracture toughness requirements can be used as a material for Containment Vessels. DPC specific considerations in standards and rules will be focused on in this paper. Additionally, comparison with the ASME Code will be discussed.

Changes in spent nuclear fuel due to dry interim storage

2012 ◽  
Vol 1475 ◽  
Author(s):  
L. Duro ◽  
O. Riba ◽  
A. Martínez-Esparza ◽  
J. Bruno
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Initial Conditions ◽  
Critical Discussion ◽  
Fuel Pellet ◽  
Spent Fuel ◽  
Geological Repository ◽  
Irradiation Period ◽  
Interim Storage ◽  
Small Bubbles

ABSTRACTThe assessment of the main changes expected for spent nuclear fuel from its discharge to its deposition in a deep geological repository is of the outmost relevance to establish the initial conditions of the disposal. In this work, a literature review and a critical discussion of the main processes that will affect the structure and the inventory of the spent nuclear fuel during its interim dry storage is presented. Once the irradiation period is finished, the following changes are observed: i) the fuel pellet is fragmented due to the temperature gradient established during the irradiation stage. On average between 10-15 fragments are observed per pellet. ii) the initial gap existing between the pellet and the cladding decreases or disappears depending on the burnup. iii) a radial zonation is observed in the microstructure of the pellet. For burnup over 40MWd/KgU, the rim develops a porosity increase due to the high local burnup and the low temperature in the periphery. The rim also presents small bubbles of fission gases. This high burnup structure implies a degradation of the thermic conductivity in the pellet, that leads to a temperature increase in the center of the pellet with a subsequent migration of the fission gases and other impurities to the grain boundaries. The implications that all these changes may have on the spent fuel behaviour is presented and discussed.

Chemical Effects at the Reaction Front in Corroding Spent Nuclear Fuel

2006 ◽  
Vol 985 ◽  
Author(s):  
Jeffrey A. Fortner ◽  
A. Jeremy Kropf ◽  
James L. Jerden ◽  
James C. Cunnane
Keyword(s):  
Transition Zone ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Aqueous Environment ◽  
Spent Fuel ◽  
Aqueous Corrosion ◽  
Reduction Potentials ◽  
Chemical Effects ◽  
Before And After ◽  
X Ray Absorption

AbstractPerformance assessment models of the U. S. repository at Yucca Mountain, Nevada suggest that neptunium from spent nuclear fuel is a potentially important dose contributor. A scientific understanding of how the UO2 matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of Np is needed to predict the behavior of Np at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within uranium oxide spent nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the Pu and Np in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO2 matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl (UO22+) mineral assemblage that is depleted in Np and Pu relative to the parent fuel. The transition from U(IV) in the fuel to a fully U(VI) character across the corrosion front is not sharp, but occurs over a transition zone of ∼ 50 micrometers. We find evidence of a thin (∼ 20 micrometer) layer that is enriched in Pu and Np within a predominantly U(IV) environment on the fuel side of the transition zone. These experimental observations are consistent with available data for the standard reduction potentials for NpO2+/Np4+ and UO22+/U4+ couples, which indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potential of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions.

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.

In-Service Inspection of Extended Dry Storage of Spent Nuclear Fuel, Part I: Crawler Technology Development

2021 ◽  
Author(s):  
Ryan M. Meyer ◽  
Jeremy Renshaw ◽  
Jamie Beard ◽  
Jon Tatman ◽  
Matt Keene ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Technology Development ◽  
Storage System ◽  
Dry Storage ◽  
Majority Population ◽  
The Usa ◽  
Extended Operation ◽  
Interim Storage ◽  
Physical Access

Abstract This paper describes development and demonstration of remote crawling systems to support periodic examinations of interim dry storage system (DSS) canisters for spent nuclear fuel in the USA. Specifically, this work relates to robotic crawler developments for “canister” based DSS systems, which form the majority population of DSSs in the USA for interim storage of spent nuclear fuel. Consideration of potential degradation of the welded stainless-steel canister in these systems is required for continued usage in the period of extended operation (PEO) beyond their initial licensed or certified terms. Challenges with performing the periodic examinations are associated with physical access to the canister surface, which is constrained due to narrow annulus spaces between the canister and the overpack, tortuous entry pathways, and high temperatures and radiation doses that can be damaging to materials and electronics. Motivations for performing periodic examinations and developing robotic crawlers for performing those examinations remotely will be presented, and several activities to demonstrate robotic crawlers for different DSS systems are summarized.

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