WorldWideScience

Sample records for aluminum-based spent nuclear

  1. Nonproliferation impacts assessment for the management of the Savannah River Site aluminum-based spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-01

    On May 13, 1996, the US established a new, 10-year policy to accept and manage foreign research reactor spent nuclear fuel containing uranium enriched in the US. The goal of this policy is to reduce civilian commerce in weapons-usable highly enriched uranium (HEU), thereby reducing the risk of nuclear weapons proliferation. Two key disposition options under consideration for managing this fuel include conventional reprocessing and new treatment and packaging technologies. The Record of Decision specified that, while evaluating the reprocessing option, ``DOE will commission or conduct an independent study of the nonproliferation and other (e.g., cost and timing) implications of chemical separation of spent nuclear fuel from foreign research reactors.`` DOE`s Office of Arms Control and Nonproliferation conducted this study consistent with the aforementioned Record of Decision. This report addresses the nonproliferation implications of the technologies under consideration for managing aluminum-based spent nuclear fuel at the Savannah River Site. Because the same technology options are being considered for the foreign research reactor and the other aluminum-based spent nuclear fuels discussed in Section ES.1, this report addresses the nonproliferation implications of managing all the Savannah River Site aluminum-based spent nuclear fuel, not just the foreign research reactor spent nuclear fuel. The combination of the environmental impact information contained in the draft EIS, public comment in response to the draft EIS, and the nonproliferation information contained in this report will enable the Department to make a sound decision regarding how to manage all aluminum-based spent nuclear fuel at the Savannah River Site.

  2. Nonproliferation impacts assessment for the management of the Savannah River Site aluminum-based spent nuclear fuel

    International Nuclear Information System (INIS)

    On May 13, 1996, the US established a new, 10-year policy to accept and manage foreign research reactor spent nuclear fuel containing uranium enriched in the US. The goal of this policy is to reduce civilian commerce in weapons-usable highly enriched uranium (HEU), thereby reducing the risk of nuclear weapons proliferation. Two key disposition options under consideration for managing this fuel include conventional reprocessing and new treatment and packaging technologies. The Record of Decision specified that, while evaluating the reprocessing option, ''DOE will commission or conduct an independent study of the nonproliferation and other (e.g., cost and timing) implications of chemical separation of spent nuclear fuel from foreign research reactors.'' DOE's Office of Arms Control and Nonproliferation conducted this study consistent with the aforementioned Record of Decision. This report addresses the nonproliferation implications of the technologies under consideration for managing aluminum-based spent nuclear fuel at the Savannah River Site. Because the same technology options are being considered for the foreign research reactor and the other aluminum-based spent nuclear fuels discussed in Section ES.1, this report addresses the nonproliferation implications of managing all the Savannah River Site aluminum-based spent nuclear fuel, not just the foreign research reactor spent nuclear fuel. The combination of the environmental impact information contained in the draft EIS, public comment in response to the draft EIS, and the nonproliferation information contained in this report will enable the Department to make a sound decision regarding how to manage all aluminum-based spent nuclear fuel at the Savannah River Site

  3. Report on the Savannah River Site aluminum-based spent nuclear fuel alternatives cost study

    International Nuclear Information System (INIS)

    Initial estimates of costs for the interim management and disposal of aluminum-based spent nuclear fuel (SNF) were developed during preparation of the Environmental Impact Statement (EIS) on the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. The Task Team evaluated multiple alternatives, assessing programmatic, technical, and schedule risks, and generated life-cycle cost projections for each alternative. The eight technology alternatives evaluated were: direct co-disposal; melt and dilute; reprocessing; press and dilute; glass material oxidation dissolution system (GMODS); electrometallurgical treatment; dissolve and vitrify; and plasma arc. In followup to the Business Plan that was developed to look at SNF dry storage, WSRC prepared an addendum to the cost study. This addendum estimated the costs for the modification and use of an existing (105L) reactor facility versus a greenfield approach for new facilities (for the Direct Co-Disposal and Melt and Dilute alternatives). WSRC assessed the impacts of a delay in reprocessing due to the potential reservation of H-Canyon for other missions (i.e., down blending HEU for commercial use or the conversion of plutonium to either MOX fuel or an immobilized repository disposal form). This report presents the relevant results from these WSRC cost studies, consistent with the most recent project policy, technology implementation, canyon utilization, and inventory assumptions. As this is a summary report, detailed information on the technical alternatives or the cost assumptions raised in each of the above-mentioned cost studies is not provided. A comparison table that briefly describes the bases used for the WSRC analyses is included as Appendix A

  4. A Multi-Attribute Utility Decision Analysis for Treatment Alternatives for the DOE/SR Aluminum-Based Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Davis, F.; Kuzio, K.; Sorenson, K.; Weiner, R.; Wheeler, T.

    1998-11-01

    A multi-attribute utility analysis is applied to the decision to select a treatment method for the management of aluminum-based spent nuclear i%el (A1-SNF) owned by the United States Department of Energy (DOE). DOE will receive, treat, and temporarily store Al- SNF, most of which is composed of highly enriched uranium, at its Savannah River Site in South Carolina. DOE intends ultimately to send the treated Al-SNJ? to a geologic repository for permanent disposal. DOE initially considered ten treatment alternatives for the management of A1-SNF, and has narrowed the choice to two of these the direct disposal and melt and dilute alternatives. The decision analysis presented in this document focuses on a decision between these two remaining alternatives.

  5. Dissolution rates of aluminum-based spent fuels relevant to geological disposal

    International Nuclear Information System (INIS)

    The Department of Energy is pursuing the option of direct disposal of a wide variety of spent nuclear fuels under its jurisdiction. Characterization of the various types of spent fuel is required prior to licensing by the Nuclear Regulatory Commission and acceptance of the fuel at a repository site. One category of required data is the expected rate of radionuclide and fissile release to the environment as a result of exposure to groundwater after closure of the repository. To provide this type of data for four different aluminum-based spent fuels, tests were conducted using a flow through method that allows the dissolution rate of the spent fuel matrix to be measured without interference by secondary precipitation reactions that would muddle interpretation of the results. Similar tests had been conducted earlier with light water reactor spent fuel, thereby allowing direct comparisons

  6. Spent nuclear fuel storage

    International Nuclear Information System (INIS)

    When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)

  7. Spent Nuclear Fuel Alternative Technology Decision Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, C.B.

    1999-11-29

    The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.

  8. Basis for Functional Performance Requirements for a Spent Nuclear Fuel Treatment and Storage Facility

    International Nuclear Information System (INIS)

    The US Department of Energy has selected the Savannah River Site (SRS) as the location to consolidate and store aluminum-based spent nuclear fuel (Al-SNF) from domestic and foreign research reactors. This report presents the technical basis for the functional performance requirements

  9. Disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste

  10. Disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste.

  11. Melt-Dilute Form of AI-Based Spent Nuclear Fuel Disposal Criticality Summary Report

    International Nuclear Information System (INIS)

    Criticality analysis of the proposed melt-dilute (MD) form of aluminum-based spent nuclear fuel (SNF), under geologic repository conditions, was performed [1] following the methodology documented in the Disposal Criticality Analysis Methodology Topical Report [2]. This methodology evaluates the potential for nuclear criticality for a waste form in a waste package. Criticality calculations show that even with waste package failure, followed by degradation of material within the waste package and potential loss of neutron absorber materials, sub-critical conditions can be readily demonstrated for the MD form of aluminum-based SNF

  12. Dry spent nuclear fuel transfer

    International Nuclear Information System (INIS)

    Newport News Shipbuilding, (NNS), has been transferring spent nuclear fuel in a dry condition for over 25 years. It is because of this successful experience that NNS decided to venture into the design, construction and operation of a commercial dry fuel transfer project. NNS is developing a remote handling system for the dry transfer of spent nuclear fuel. The dry fuel transfer system is applicable to spent fuel pool-to-cask or cask-to-cask or both operations. It is designed to be compatible with existing storage cask technology as well as the developing multi-purpose canister design. The basis of NNS' design is simple. It must be capable of transferring all fuel designs, it must be capable of servicing 100 percent of the commercial nuclear plants, it must protect the public and nuclear operators, it must be operated cost efficiently and it must be transportable. Considering the basic design parameters, the following are more specific requirements included in the design: (a) Total weight of transfer cask less than 24 tons; (b) no requirement for permanent site modifications to support system utilization; (c) minimal radiation dose to operating personnel; (d) minimal generation of radioactive waste; (e) adaptability to any size and length fuel or cask; (f) portability of system allowing its efficient movement from site to site; (g) safe system; all possible ''off normal'' situations are being considered, and resultant safety systems are being engineered into NNS' design to mitigate problems. The primary focus of this presentation is to provide an overview of NNS' Dry Spent Nuclear Fuel Transfer System. (author). 5 refs

  13. Antineutrino monitoring of spent nuclear fuel

    OpenAIRE

    Brdar, Vedran; Huber, Patrick; Kopp, Joachim

    2016-01-01

    Military and civilian applications of nuclear energy have left a significant amount of spent nuclear fuel over the past 70 years. Currently, in many countries world wide, the use of nuclear energy is on the rise. Therefore, the management of highly radioactive nuclear waste is a pressing issue. In this letter, we explore antineutrino detectors as a tool for monitoring and safeguarding nuclear waste material. We compute the flux and spectrum of antineutrinos emitted by spent nuclear fuel eleme...

  14. Spent nuclear fuel reprocessing modeling

    International Nuclear Information System (INIS)

    The long-term wide development of nuclear power requires new approaches towards the realization of nuclear fuel cycle, namely, closed nuclear fuel cycle (CNFC) with respect to fission materials. Plant nuclear fuel cycle (PNFC), which is in fact the reprocessing of spent nuclear fuel unloaded from the reactor and the production of new nuclear fuel (NF) at the same place together with reactor plant, can be one variant of CNFC. Developing and projecting of PNFC is a complicated high-technology innovative process that requires modern information support. One of the components of this information support is developed by the authors. This component is the programme conducting calculations for various variants of process flow sheets for reprocessing SNF and production of NF. Central in this programme is the blocks library, where the blocks contain mathematical description of separate processes and operations. The calculating programme itself has such a structure that one can configure the complex of blocks and correlations between blocks, appropriate for any given flow sheet. For the ready sequence of operations balance calculations are made of all flows, i.e. expenses, element and substance makeup, heat emission and radiation rate are determined. The programme is open and the block library can be updated. This means that more complicated and detailed models of technological processes will be added to the library basing on the results of testing processes using real equipment, in test operating mode. The development of the model for the realization of technical-economic analysis of various variants of technologic PNFC schemes and the organization of 'operator's advisor' is expected. (authors)

  15. Safety aspects of receipt and storage of spent nuclear fuel at the Savannah River site

    Energy Technology Data Exchange (ETDEWEB)

    Busby, A.S.; Andes, T.C. [Westinghouse Savannah River Co., Aiken, SC 29808 (United States)

    2001-07-01

    The Savannah River Site receives and stores aluminum-based spent nuclear fuel from research reactors world-wide in support of the U.S. Department of Energy's take back policy for U.S. origin enriched uranium. For over 35 years the Savannah River Site has supported this policy in a safe and deliberate manner. Facilities dedicated to this mission include the Receiving Basin for Offsite Fuels and L-Basin. Current inventories are about 6,500 aluminum-based research reactor assemblies and about 700 stainless steel or zirconium clad prototype power reactor assemblies. The purpose of this paper is to briefly describe the processes that the Savannah River Site employs to safely receive, handle, and store spent nuclear fuel. (author)

  16. Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement

    Energy Technology Data Exchange (ETDEWEB)

    N/A

    2000-04-14

    The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel 20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign and domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some Americium/Curium Targets stored at SRS. Alternatives considered in this EIS encompass a range of new packaging, new processing, and conventional processing technologies, as well as the No Action Alternative. A preferred alternative is identified in which DOE would prepare about 97% by volume (about 60% by mass) of the aluminum-based fuel for disposition using a melt and dilute treatment process. The remaining 3% by volume (about 40% by mass) would be managed using chemical separation. Impacts are assessed primarily in the areas of water resources, air resources, public and worker health, waste management, socioeconomic, and cumulative impacts.

  17. Spent Nuclear Fuel (SNF) Project Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    LEROY, P.G.

    2000-11-03

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

  18. Spent Nuclear Fuel (SNF) Project Execution Plan

    International Nuclear Information System (INIS)

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities

  19. Investigation of Spent Nuclear Fuel Pool Coolability

    OpenAIRE

    Nimander, Fredrik

    2011-01-01

    The natural catastrophe at Fukushima Dai-ichi 2011 enlightened the nuclear community. This master thesis reveals the non-negligible risks regarding the short term storage of spent nuclear fuel. The thesis has also investigated the possibility of using natural circulation of air in a passive safety system to cool the spent nuclear fuel pools. The results where conclusive: The temperature difference between the heated air and ambient air is far too low for natural circulation of air to remove a...

  20. Spent nuclear fuel disposal liability insurance

    International Nuclear Information System (INIS)

    This thesis examines the social efficiency of nuclear power when the risks of accidental releases of spent fuel radionuclides from a spent fuel disposal facility are considered. The analysis consists of two major parts. First, a theoretical economic model of the use of nuclear power including the risks associated with releases of radionuclides from a disposal facility is developed. Second, the costs of nuclear power, including the risks associated with a radionuclide release, are empirically compared to the costs of fossil fuel-fired generation of electricity. Under the provisions of the Nuclear Waste Policy Act of 1982, the federally owned and operated spent nuclear fuel disposal facility is not required to maintain a reserve fund to cover damages from an accidental radionuclide release. Thus, the risks of a harmful radionuclide release are not included in the spent nuclear fuel disposal fee charged to the electric utilities. Since the electric utilities do not pay the full, social costs of spent fuel disposal, they use nuclear fuel in excess of the social optimum. An insurance mechanism is proposed to internalize the risks associated with spent fueled disposal. Under this proposal, the Federal government is required to insure the disposal facility against any liabilities arising from accidental releases of spent fuel radionuclides

  1. Nuclear spent fuel management. Experience and options

    International Nuclear Information System (INIS)

    Spent nuclear fuel can be stored safely for long periods at relatively low cost, but some form of permanent disposal will eventually be necessary. This report examines the options for spent fuel management, explores the future prospects for each stage of the back-end of the fuel cycle and provides a thorough review of past experience and the technical status of the alternatives. Current policies and practices in twelve OECD countries are surveyed

  2. Nondestructive measurements on spent fuel for the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Nondestructive measurements on spent fuel are being developed to meet safeguards and materials managment requirements at nuclear facilities. Spent-fuel measurement technology and its applications are reviewed

  3. Integrated spent nuclear fuel database system

    International Nuclear Information System (INIS)

    The Distributed Information Systems software Unit at the Idaho National Engineering Laboratory has designed and developed an Integrated Spent Nuclear Fuel Database System (ISNFDS), which maintains a computerized inventory of all US Department of Energy (DOE) spent nuclear fuel (SNF). Commercial SNF is not included in the ISNFDS unless it is owned or stored by DOE. The ISNFDS is an integrated, single data source containing accurate, traceable, and consistent data and provides extensive data for each fuel, extensive facility data for every facility, and numerous data reports and queries

  4. Long term wet spent nuclear fuel storage

    International Nuclear Information System (INIS)

    The meeting showed that there is continuing confidence in the use of wet storage for spent nuclear fuel and that long-term wet storage of fuel clad in zirconium alloys can be readily achieved. The importance of maintaining good water chemistry has been identified. The long-term wet storage behaviour of sensitized stainless steel clad fuel involves, as yet, some uncertainties. However, great reliance will be placed on long-term wet storage of spent fuel into the future. The following topics were treated to some extent: Oxidation of the external surface of fuel clad, rod consolidation, radiation protection, optimum methods of treating spent fuel storage water, physical radiation effects, and the behaviour of spent fuel assemblies of long-term wet storage conditions. A number of papers on national experience are included

  5. Spent Nuclear Fuel Storage Program user's guide

    International Nuclear Information System (INIS)

    The purpose of this manual is to present procedures to execute the Spent Nuclear Fuel Storage Model (SNFSM) program. This manual includes an overview of the model, operating environment, input and output specifications and user procedures. An example of the execution of the program is included to assist potential users

  6. Classification of spent nuclear fuel (SNF)

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high-priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. This document discusses the classification of spent nuclear fuels

  7. A present status for dry storage of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bang, K. S.; Lee, J. C.; Park, H. Y.; Seo, K. S

    2003-04-01

    National policy for management of a spent nuclear fuel does not establish in Korea yet. A storage capacity of a storage pool that is to store the spent nuclear fuel will be exceeded an amount of accumulation from the first Woljin nuclear power plant in 2007. Therefore it is necessary that dry storage facility is secured to store safely the spent nuclear fuel on site of the nuclear power plant until national policy for a back-end spent nuclear fuel cycle is established. In order to store safely spent nuclear fuel, it is important that the present status and technology on dry storage of spent nuclear fuel is looked over. Therefore, the present status on dry storage of spent nuclear fuel was analyzed so as to develop dry storage system and choose a proper dry storage method domestic.

  8. Antineutrino monitoring of spent nuclear fuel

    CERN Document Server

    Brdar, Vedran; Kopp, Joachim

    2016-01-01

    Military and civilian applications of nuclear energy have left a significant amount of spent nuclear fuel over the past 70 years. Currently, in many countries world wide, the use of nuclear energy is on the rise. Therefore, the management of highly radioactive nuclear waste is a pressing issue. In this letter, we explore antineutrino detectors as a tool for monitoring and safeguarding nuclear waste material. We compute the flux and spectrum of antineutrinos emitted by spent nuclear fuel elements as a function of time, and we illustrate the usefulness of antineutrino detectors in several benchmark scenarios. In particular, we demonstrate how a measurement of the antineutrino flux can help to re-verify the contents of a dry storage cask in case the monitoring chain by conventional means gets disrupted. We then comment on the usefulness of antineutrino detectors at long-term storage facilities such as Yucca mountain. Finally, we put forward antineutrino detection as a tool in locating underground "hot spots" in ...

  9. Management and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    The programme consists of the long-term and short-term programme, the continued bedrock investigations, the underground research laboratory, the decision-making procedure in the site selection process and information questions during the site selection process. The National Board for Spent Nuclear Fuel hereby subunits both the SKB's R and D Programme 86 and the Board's statement concerning the programme. Decisions in the matter have been made by the Board's executive committee. (DG)

  10. Spent nuclear fuel project integrated schedule plan

    Energy Technology Data Exchange (ETDEWEB)

    Squires, K.G.

    1995-03-06

    The Spent Nuclear Fuel Integrated Schedule Plan establishes the organizational responsibilities, rules for developing, maintain and status of the SNF integrated schedule, and an implementation plan for the integrated schedule. The mission of the SNFP on the Hanford site is to provide safe, economic, environmentally sound management of Hanford SNF in a manner which stages it to final disposition. This particularly involves K Basin fuel.

  11. Report on interim storage of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  12. Nuclear Spent Fuel Management in Spain

    International Nuclear Information System (INIS)

    The radioactive waste management policy is established by the Spanish Government through the Ministry of Industry, Tourism and Commerce. This policy is described in the Cabinet-approved General Radioactive Waste Plan. ENRESA is the Spanish organization in charge of radioactive waste and nuclear SFM and nuclear installations decommissioning. The priority goal in SFM is the construction of the centralized storage facility named Almacén Temporal Centralizado (ATC), whose generic design was approved by the safety authority, Consejo de Seguridad Nuclear. This facility is planned for some 6.700 tons of heavy metal. The ATC site selection process, based on a volunteer community’s scheme, has been launched by the Government in December 2009. After the selection of a site in a participative and transparent process, the site characterization and licensing activities will support the construction of the facility. Meanwhile, extension of the on-site storage capacity has been implemented at the seven nuclear power plants sites, including past reracking at all sites. More recent activities are: reracking performed at Cofrentes NPP; dual purpose casks re-licensing for higher burnup at Trillo NPP; transfer of the spent fuel inventory at Jose Cabrera NPP to a dry-storage system, to allow decommissioning operations; and licence application of a dry-storage installation at Ascó NPP, to provide the needed capacity until the ATC facility operation. For financing planning purposes, the long-term management of spent fuel is based on direct disposal. A final decision about major fuel management options is not made yet. To assist the decision makers a number of activities are under way, including basic designs of a geological disposal facility for clay and granite host rocks, together with associated performance assessment, and supported by a R&D programme, which also includes research projects in other options like advanced separation and transmutation. (author)

  13. Transportation capabilities study of DOE-owned spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Clark, G.L.; Johnson, R.A.; Smith, R.W. [Packaging Technology, Inc., Tacoma, WA (United States); Abbott, D.G.; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1994-10-01

    This study evaluates current capabilities for transporting spent nuclear fuel owned by the US Department of Energy. Currently licensed irradiated fuel shipping packages that have the potential for shipping the spent nuclear fuel are identified and then matched against the various spent nuclear fuel types. Also included are the results of a limited investigation into other certified packages and new packages currently under development. This study is intended to support top-level planning for the disposition of the Department of Energy`s spent nuclear fuel inventory.

  14. Spent Nuclear Fuel Vibration Integrity Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Jiang, Hao [ORNL; Yan, Yong [ORNL; Bevard, Bruce Balkcom [ORNL

    2016-01-01

    The objective of this research is to collect dynamic experimental data on spent nuclear fuel (SNF) under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT), the hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). The collected CIRFT data will be utilized to support ongoing spent fuel modeling activities, and support SNF transportation related licensing issues. Recent testing to understand the effects of hydride reorientation on SNF vibration integrity is also being evaluated. CIRFT results have provided insight into the fuel/clad system response to transportation related loads. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance, Fuel structure contributes to the SNF system stiffness, There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interaction, and SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous. Because of the non-homogeneous composite structure of the SNF system, finite element analyses (FEA) are needed to translate the global moment-curvature measurement into local stress-strain profiles. The detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained directly from a CIRFT system measurement. Therefore, detailed FEA is used to understand the global test response, and that data will also be presented.

  15. Overview of the United States spent nuclear fuel program

    International Nuclear Information System (INIS)

    As a result of the end of the Cold War, the mission of the US Department of Energy (DOE) has shifted from an emphasis on nuclear weapons development and production to an emphasis on the safe management and disposal of excess nuclear materials including spent nuclear fuel from both production and research reactors. Within the US, there are two groups managing spent nuclear fuel. Commercial nuclear power plants are managing their spent nuclear fuel at the individual reactor sites until the planned repository is opened. All other spent nuclear fuel, including research reactors, university reactors, naval reactors, and legacy material from the Cold War is managed by DOE. DOE's mission is to safely and efficiently manage its spent nuclear fuel and prepare it for disposal. This mission involves correcting existing vulnerabilities in spent fuel storage; moving spent fuel from wet basins to dry storage; processing at-risk spent fuel; and preparing spent fuel in road-ready condition for repository disposal. Most of DOE's spent nuclear fuel is stored in underwater basins (wet storage). Many of these basins are outdated, and spent fuel is to be removed and transferred to more modern basins or to new dry storage facilities. In 1995, DOE completed a complex-wide environmental impact analysis that resulted in spent fuel being sent to one of three principal DOE sites for interim storage (up to 40 years) prior to shipment to a repository. This regionalization by fuel type will allow for economies of scale yet minimize unnecessary transportation. This paper discusses the national SNF program, ultimate disposition of SNF, and the technical challenges that have yet to be resolved, namely, release rate testing, non-destructive assay, alternative treatments, drying, and chemical reactivity

  16. Advanced waste forms from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ackerman, J.P.; McPheeters, C.C.

    1995-12-31

    More than one hundred spent nuclear fuel types, having an aggregate mass of more than 5000 metric tons (2700 metric tons of heavy metal), are stored by the United States Department of Energy. This paper proposes a method for converting this wide variety of fuel types into two waste forms for geologic disposal. The method is based on a molten salt electrorefining technique that was developed for conditioning the sodium-bonded, metallic fuel from the Experimental Breeder Reactor-II (EBR-II) for geologic disposal. The electrorefining method produces two stable, optionally actinide-free, high-level waste forms: an alloy formed from stainless steel, zirconium, and noble metal fission products, and a ceramic waste form containing the reactive metal fission products. Electrorefining and its accompanying head-end process are briefly described, and methods for isolating fission products and fabricating waste forms are discussed.

  17. Onsite storage of spent nuclear fuel in metalic spent fuel storage casks

    International Nuclear Information System (INIS)

    Virginia Electric and Power Company (Vepco) owns and operates two nuclear power stations within its system: the North Anna Power Station located in Louisa County, Virginia; and the Surry Power Station located in Surry County, Virginia. Each of these power stations has two pressurized water reactor operating units which share a common spent fuel pool. Under the Nuclear Waste Policy Act of 1982, Vepco is responsible for providing interim spent fuel storage until availability of the Federal Repository. Vepco has studied a number of options and has developed a program to provide the required onsite interim spent fuel storage. Options considered by Vepco included reracking, pin consolidation, dry storage and construction of a new spent fuel pool to provide the increased spent fuel storage capacity required. Vepco has selected reracking at North Anna combined with dry storage in metal spent fuel storage casks at Surrey to provide the required onsite spent fuel storage. A dry cask storage facility design and license application were developed and the license application was submitted to the NRC in October, 1982. The selection of the option to use dry cask storage of spent fuel at Surry represents the first attempt to license dry storage of spent nuclear fuel in the United States. This storage option is expected to provide an effective option for utilities without adequate storage space in their existing spent fuel pools

  18. Spent nuclear fuel project quality assurance program plan

    Energy Technology Data Exchange (ETDEWEB)

    Lacey, R.E.

    1997-05-09

    This main body of this document describes how the requirements of 10 CFR 830.120 are met by the Spent Nuclear Fuel Project through implementation of WHC-SP-1131. Appendix A describes how the requirements of DOE/RW-0333P are met by the Spent Nuclear Fuel Project through implementation of specific policies, manuals, and procedures.

  19. Alternative measuring approaches in gamma scanning on spent nuclear fuel

    OpenAIRE

    Sihm Kvenangen, Karen

    2007-01-01

    In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific ga...

  20. Site selection - location of the repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    This document describes the localization work and SKB's choice of site for the repository. Furthermore, SKB's basis and rationale for the decisions taken during the work are reported. The document is Appendix PV of applications under the Nuclear Activities Act and the Environmental Code to both build and operate an encapsulation plant adjacent to the central interim storage facility for spent nuclear fuel in Oskarshamn, and to construct and operate a disposal facility for spent nuclear fuel at Forsmark in Oesthammar municipality

  1. Near-field chemistry of the spent nuclear fuel repository

    International Nuclear Information System (INIS)

    Factors affecting near-field chemistry of the spent nuclear fuel repository as well as the involved mutual interactions are described on the basis of literature. The most important processes in the near-field (spent-fuel, canister and bentonite) are presented. The related examples on near-field chemistry models shed light on the extensive problematics of near-field chemistry. (authors)

  2. Microbiology of spent nuclear fuel storage basins.

    Science.gov (United States)

    Santo Domingo, J W; Berry, C J; Summer, M; Fliermans, C B

    1998-12-01

    Microbiological studies of spent nuclear fuel storage basins at Savannah River Site (SRS) were performed as a preliminary step to elucidate the potential for microbial-influenced corrosion (MIC) in these facilities. Total direct counts and culturable counts performed during a 2-year period indicated microbial densities of 10(4) to 10(7) cells/ml in water samples and on submerged metal coupons collected from these basins. Bacterial communities present in the basin transformed between 15% and 89% of the compounds present in Biologtrade mark plates. Additionally, the presence of several biocorrosion-relevant microbial groups (i.e., sulfate-reducing bacteria and acid-producing bacteria) was detected with commercially available test kits. Scanning electron microscopy and X-ray spectra analysis of osmium tetroxide-stained coupons demonstrated the development of microbial biofilm communities on some metal coupons submerged for 3 weeks in storage basins. After 12 months, coupons were fully covered by biofilms, with some deterioration of the coupon surface evident at the microscopical level. These results suggest that, despite the oligotrophic and radiological environment of the SRS storage basins and the active water deionization treatments commonly applied to prevent electrochemical corrosion in these facilities, these conditions do not prevent microbial colonization and survival. Such microbial densities and wide diversity of carbon source utilization reflect the ability of the microbial populations to adapt to these environments. The presumptive presence of sulfate-reducing bacteria and acid-producing bacteria and the development of biofilms on submerged coupons indicated that an environment for MIC of metal components in the storage basins may occur. However, to date, there has been no indication or evidence of MIC in the basins. Basin chemistry control and corrosion surveillance programs instituted several years ago have substantially abated all corrosion mechanisms.

  3. Microbiology of spent nuclear fuel storage basins.

    Science.gov (United States)

    Santo Domingo, J W; Berry, C J; Summer, M; Fliermans, C B

    1998-12-01

    Microbiological studies of spent nuclear fuel storage basins at Savannah River Site (SRS) were performed as a preliminary step to elucidate the potential for microbial-influenced corrosion (MIC) in these facilities. Total direct counts and culturable counts performed during a 2-year period indicated microbial densities of 10(4) to 10(7) cells/ml in water samples and on submerged metal coupons collected from these basins. Bacterial communities present in the basin transformed between 15% and 89% of the compounds present in Biologtrade mark plates. Additionally, the presence of several biocorrosion-relevant microbial groups (i.e., sulfate-reducing bacteria and acid-producing bacteria) was detected with commercially available test kits. Scanning electron microscopy and X-ray spectra analysis of osmium tetroxide-stained coupons demonstrated the development of microbial biofilm communities on some metal coupons submerged for 3 weeks in storage basins. After 12 months, coupons were fully covered by biofilms, with some deterioration of the coupon surface evident at the microscopical level. These results suggest that, despite the oligotrophic and radiological environment of the SRS storage basins and the active water deionization treatments commonly applied to prevent electrochemical corrosion in these facilities, these conditions do not prevent microbial colonization and survival. Such microbial densities and wide diversity of carbon source utilization reflect the ability of the microbial populations to adapt to these environments. The presumptive presence of sulfate-reducing bacteria and acid-producing bacteria and the development of biofilms on submerged coupons indicated that an environment for MIC of metal components in the storage basins may occur. However, to date, there has been no indication or evidence of MIC in the basins. Basin chemistry control and corrosion surveillance programs instituted several years ago have substantially abated all corrosion mechanisms

  4. Savannah River Site, spent nuclear fuel management, draft environmental impact statement: Summary

    International Nuclear Information System (INIS)

    The proposed DOE action described in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets assigned to the Savannah River Site (SRS), including placing these materials informs suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel. Alternatives considered in this EIS encompass a range of new packaging, new processing, and conventional reprocessing technologies, as well as the No Action Alternative. A preferred alternative is identified in which DOE's preference is to prepare about 97% by volume (about 60% by mass) of the aluminum-based fuel for disposition using a melt and dilute treatment process. The remaining 3% by volume (about 40% by mass) would be managed using conventional processing. Impacts are assessed primarily in the areas of water resources, air resources, public and worker health, waste management, socioeconomic, and cumulative impacts

  5. Spent nuclear fuel discharges from U.S. reactors 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  6. Spent nuclear fuel discharges from U.S. reactors 1994

    International Nuclear Information System (INIS)

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year's report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs

  7. Storage of Spent Nuclear Fuel. Specific Safety Guide

    International Nuclear Information System (INIS)

    This Safety Guide provides recommendations and guidance on the storage of spent nuclear fuel. It covers all types of storage facilities and all types of spent fuel from nuclear power plants and research reactors. It takes into consideration the longer storage periods that have become necessary owing to delays in the development of disposal facilities and the decrease in reprocessing activities. It also considers developments associated with nuclear fuel, such as higher enrichment, mixed oxide fuels and higher burnup. The Safety Guide is not intended to cover the storage of spent fuel if this is part of the operation of a nuclear power plant or spent fuel reprocessing facility. Guidance is provided on all stages for spent fuel storage facilities, from planning through siting and design to operation and decommissioning, and in particular retrieval of spent fuel. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. Management system; 5. Safety case and safety assessment; 6. General safety considerations for storage of spent fuel. Appendix I: Specific safety considerations for wet or dry storage of spent fuel; Appendix II: Conditions for specific types of fuel and additional considerations; Annex: I: Short term and long term storage; Annex II: Operational and safety considerations for wet and dry spent fuel storage facilities; Annex III: Examples of sections of operating procedures for a spent fuel storage facility; Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex VI: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex VII: Postulated initiating events for consideration in a safety assessment (internal phenomena).

  8. Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay

    Energy Technology Data Exchange (ETDEWEB)

    Quiter, Brian; Ludewigt, Bernhard; Ambers, Scott

    2011-06-30

    In nuclear resonance fluorescence (NRF) measurements, resonances are excited by an external photon beam leading to the emission of gamma rays with specific energies that are characteristic of the emitting isotope. NRF promises the unique capability of directly quantifying a specific isotope without the need for unfolding the combined responses of several fissile isotopes as is required in other measurement techniques. We have analyzed the potential of NRF as a non-destructive analysis technique for quantitative measurements of Pu isotopes in spent nuclear fuel (SNF). Given the low concentrations of 239Pu in SNF and its small integrated NRF cross sections, the main challenge in achieving precise and accurate measurements lies in accruing sufficient counting statistics in a reasonable measurement time. Using analytical modeling, and simulations with the radiation transport code MCNPX that has been experimentally tested recently, the backscatter and transmission methods were quantitatively studied for differing photon sources and radiation detector types. Resonant photon count rates and measurement times were estimated for a range of photon source and detection parameters, which were used to determine photon source and gamma-ray detector requirements. The results indicate that systems based on a bremsstrahlung source and present detector technology are not practical for high-precision measurements of 239Pu in SNF. Measurements that achieve the desired uncertainties within hour-long measurements will either require stronger resonances, which may be expressed by other Pu isotopes, or require quasi-monoenergetic photon sources with intensities that are approximately two orders of magnitude higher than those currently being designed or proposed.This work is part of a larger effort sponsored by the Next Generation Safeguards Initiative to develop an integrated instrument, comprised of individual NDA techniques with complementary features, that is fully capable of

  9. Technical bases for interim storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    The experience base for water storage of spent nuclear fuel has evolved since 1943. The technology base includes licensing documentation, standards, technology studies, pool operator experience, and documentation from public hearings. That base reflects a technology which is largely successful and mundane. It projects probable satisfactory water storage of spent water reactor fuel for several decades. Interim dry storage of spent water reactor fuel is not yet licensed in the US, but a data base and documentation have developed. There do not appear to be technological barriers to interim dry storage, based on demonstrations with irradiated fuel. Water storage will continue to be a part of spent fuel management at reactors. Whether dry storage becomes a prominent interim fuel management option depends on licensing and economic considerations. National policies will strongly influence how long the spent fuel remains in interim storage and what its final disposition will be

  10. Separator assembly for use in spent nuclear fuel shipping cask

    Science.gov (United States)

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  11. Spent nuclear fuel discharges from US reactors 1993

    Energy Technology Data Exchange (ETDEWEB)

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  12. Disposal of spent nuclear fuel in geological repositories

    International Nuclear Information System (INIS)

    Plutonium has been and still is produced in the world's many civil and military nuclear programs. Although the nuclear establishments of several countries, most noticeably, Japan, France, Greta Britain and Russia, are advocating the recycling of plutonium, there are also two nuclear waste disposition 'strategies' that involve the direct final disposal of plutonium in geological repositories: the direct final disposal of spent fuel or spent civil mixed oxide fuel when option to reprocess it has been rejected; and the final disposal of 'spent fuel standard' excess weapon plutonium when it has been 'anti-reprocessed' or burned as 'military mixed oxide fuel. It is important to understand that the magnitude of long-term safeguards concern of plutonium disposal in geological repositories depends very much on the future development of nuclear energy application. It might be decided to solve the global plutonium predicament by making global stocks of plutonium 'irretrievable', thus removing the needs for safeguards

  13. Separation of actinides from spent nuclear fuel: A review.

    Science.gov (United States)

    Veliscek-Carolan, Jessica

    2016-11-15

    This review summarises the methods currently available to extract radioactive actinide elements from solutions of spent nuclear fuel. This separation of actinides reduces the hazards associated with spent nuclear fuel, such as its radiotoxicity, volume and the amount of time required for its' radioactivity to return to naturally occurring levels. Separation of actinides from environmental water systems is also briefly discussed. The actinide elements typically found in spent nuclear fuel include uranium, plutonium and the minor actinides (americium, neptunium and curium). Separation methods for uranium and plutonium are reasonably well established. On the other hand separation of the minor actinides from lanthanide fission products also present in spent nuclear fuel is an ongoing challenge and an area of active research. Several separation methods for selective removal of these actinides from spent nuclear fuel will be described. These separation methods include solvent extraction, which is the most commonly used method for radiochemical separations, as well as the less developed but promising use of adsorption and ion-exchange materials.

  14. Separation of actinides from spent nuclear fuel: A review.

    Science.gov (United States)

    Veliscek-Carolan, Jessica

    2016-11-15

    This review summarises the methods currently available to extract radioactive actinide elements from solutions of spent nuclear fuel. This separation of actinides reduces the hazards associated with spent nuclear fuel, such as its radiotoxicity, volume and the amount of time required for its' radioactivity to return to naturally occurring levels. Separation of actinides from environmental water systems is also briefly discussed. The actinide elements typically found in spent nuclear fuel include uranium, plutonium and the minor actinides (americium, neptunium and curium). Separation methods for uranium and plutonium are reasonably well established. On the other hand separation of the minor actinides from lanthanide fission products also present in spent nuclear fuel is an ongoing challenge and an area of active research. Several separation methods for selective removal of these actinides from spent nuclear fuel will be described. These separation methods include solvent extraction, which is the most commonly used method for radiochemical separations, as well as the less developed but promising use of adsorption and ion-exchange materials. PMID:27427893

  15. Savannah River Site, spent nuclear fuel management, draft environmental impact statement

    International Nuclear Information System (INIS)

    The management of spent nuclear fuel (SNF) has been an integral part of the mission of the Savannah River Site (SRS) for more than 40 years. Until the early 1990s, SNF management consisted primarily of short-term onsite storage and reprocessing in the SRS chemical separation facilities to produce strategic nuclear materials. With the end of the Cold War, the US Department of Energy (DOE) decided to phase out reprocessing of SNF for the production of nuclear weapons materials. Therefore, the management strategy for this fuel has shifted from short-term storage and reprocessing for the recovery of highly-enriched uranium and transuranic isotopes to stabilization, when necessary, and interim storage pending final disposition that includes preparing aluminum-based SNF for placement in a geologic repository. In addition to the fuel already onsite, the SRS will receive SNF from foreign research reactors until 2009 and from domestic research reactors until, potentially, 2035. As a result, the safe and efficient management of SNF will continue to be an important SRS mission. This EIS evaluates the potential environmental impacts of DOE's proposed plans for management SNF assigned to SRS

  16. Spent nuclear fuel Canister Storage Building CDR Review Committee report

    International Nuclear Information System (INIS)

    The Canister Storage Building (CSB) is a subproject under the Spent Nuclear Fuels Major System Acquisition. This subproject is necessary to design and construct a facility capable of providing dry storage of repackaged spent fuels received from K Basins. The CSB project completed a Conceptual Design Report (CDR) implementing current project requirements. A Design Review Committee was established to review the CDR. This document is the final report summarizing that review

  17. Spent nuclear fuel Canister Storage Building CDR Review Committee report

    Energy Technology Data Exchange (ETDEWEB)

    Dana, W.P.

    1995-12-01

    The Canister Storage Building (CSB) is a subproject under the Spent Nuclear Fuels Major System Acquisition. This subproject is necessary to design and construct a facility capable of providing dry storage of repackaged spent fuels received from K Basins. The CSB project completed a Conceptual Design Report (CDR) implementing current project requirements. A Design Review Committee was established to review the CDR. This document is the final report summarizing that review

  18. Public Opinion Surveys in Spent Nuclear Fuel Management

    OpenAIRE

    Vasilieva, E.

    2002-01-01

    Russia's plans to import foreign SNF for storage and reprocessing meet serious public opposition. As a start of taking into account public concerns, programs of public involvement can be designed and implemented. In the paper, approaches to decision-making on spent nuclear fuel management that differ in their commitment to public participation are discussed. The review of public opinion surveys in Russia that investigated public attitudes to spent fuel is given. Finally, the experience of sev...

  19. Damage in spent nuclear fuel defined by properties and requirements

    International Nuclear Information System (INIS)

    The Nuclear Regulatory Commission's (NRC's) Spent Fuel Program Office (SFPO) has provided guidance in defining damaged fuel in Interim Staff Guidance, ISG-1. This guidance is similar to that developed by the American National Standards Institute (ANSI). Neither of these documents gives the logic behind its definition of damaged fuel. The paper discusses the requirements placed on spent fuel for dry interim storage and transport and the ways in which service requirements drive the definition of damage for spent fuel. Examples are given to illustrate the methodology, which focuses on defining damaged fuel based on the properties that the fuel must exhibit to meet the requirements of storage and/or transport. (author)

  20. Direct reuse of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M.A., E-mail: mnader73@yahoo.com

    2014-10-15

    Highlights: • A new design for the PWR assemblies for direct use of spent fuel was proposed. • The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors. • The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. • MCNPX is used for the calculations that showed that the burnup can be increased by about 25%. • Acceptable linear heat generation rate in hot rods and improved Pu proliferation resistance. - Abstract: In this paper we proposed a new design for the PWR fuel assembly for direct use of the PWR spent fuel without processing. The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors which preferably built in the same site to avoid the problem of transportations. The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. Each tube has the same inner diameter of that of CANDU pressure tube. The spaces between the tubes contain low enriched UO{sub 2} fuel rods and guide tubes. MCNPX code is used for the simulation and calculation of the burnup of the proposed assembly. The bundles after the discharge from the PWR with their materials inventories are burned in a CANDU cell after a certain decay time. The results were compared with reference results and the impact of this new design on the uranium utilization improvement and on the proliferation resistance of plutonium is discussed. The effect of this new design on the power peaking, moderator temperature coefficient of reactivity and CANDU coolant void reactivity are discussed as well.

  1. Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography

    OpenAIRE

    Jonkmans, G.; Anghel, V. N. P.; Jewett, C.; Thompson, M.

    2012-01-01

    This paper explores the use of cosmic ray muons to image the contents of shielded containers and detect high-Z special nuclear materials inside them. Cosmic ray muons are a naturally occurring form of radiation, are highly penetrating and exhibit large scattering angles on high Z materials. Specifically, we investigated how radiographic and tomographic techniques can be effective for non-invasive nuclear waste characterization and for nuclear material accountancy of spent fuel inside dry stor...

  2. Microbial biofilm growth on irradiated, spent nuclear fuel cladding

    International Nuclear Information System (INIS)

    A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 x 103 Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments

  3. Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography

    CERN Document Server

    Jonkmans, G; Jewett, C; Thompson, M

    2012-01-01

    This paper explores the use of cosmic ray muons to image the contents of shielded containers and detect high-Z special nuclear materials inside them. Cosmic ray muons are a naturally occurring form of radiation, are highly penetrating and exhibit large scattering angles on high Z materials. Specifically, we investigated how radiographic and tomographic techniques can be effective for non-invasive nuclear waste characterization and for nuclear material accountancy of spent fuel inside dry storage containers. We show that the tracking of individual muons, as they enter and exit a structure, can potentially improve the accuracy and availability of data on nuclear waste and the contents of Dry Storage Containers (DSC) used for spent fuel storage at CANDU plants. This could be achieved in near real time, with the potential for unattended and remotely monitored operations. We show that the expected sensitivity, in the case of the DSC, exceeds the IAEA detection target for nuclear material accountancy.

  4. Nuclear reactor spent fuel valuation: procedure, applications, and analysis

    International Nuclear Information System (INIS)

    A preliminary approach that values nuclear reactor generated spent fuel is developed and applied in this report. There is no intent to assess the merits of reprocessing and recycling but rather to outline a procedure that may provide a basis for international negotiations on nation-to-nation spent fuel transfer. The valuation procedure described estimates the net present discounted value (PDV) of the benefit incurred when the reprocessed plutonium and uranium contained in the spent fuel are recycled, less the PDV of the stream of costs associated with transporting, storing, and reprocessing the spent fuel and fabricating, storing, and safeguarding the new mixed oxide fuel (MOX). The parameters that affect the net PDV most strongly are the discount rate, yellowcake prices, reprocessing costs, timing of recycle or disposal, and, to a lesser extent, enrichment costs

  5. Alternative Measuring Approaches in Gamma Scanning on Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sihm Kvenangen, Karen

    2007-06-15

    In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific gamma ray spectroscopy method called gamma scanning the gamma radiation from the fission products Cs-137, Cs-134 and Eu-154 are measured in a spent fuel assembly. From the results, conclusions can be drawn about the fuels characteristics. This degree project examines the possibilities of using alternative measuring approaches when using the gamma scanning method. The focus is on examining how to increase the quality of the measured data. How to decrease the measuring time as compared with the present measuring strategy, has also been investigated. The main part of the study comprises computer simulations of gamma scanning measurements. The simulations have been validated with actual measurements on spent nuclear fuel at the central interim storage, Clab. The results show that concerning the quality of the measuring data the conventional strategy is preferable, but with other starting positions and with a more optimized equipment. When focusing on the time aspect, the helical measuring strategy can be an option, but this needs further investigation.

  6. A GIS based methodology for nuclear reactor spent fuel disposal

    International Nuclear Information System (INIS)

    This article aims at studying the use of Geographical Information Systems for selecting a site for radioactive waste disposal of spent fuel generated by operation of Angra 1 and 2 nuclear power stations, in order to provide additional means for solving this problem in Brazil. This spent fuel continues to generate decay heat and radiation after its use in power stations. The disposal should be done in such a way as to isolate the nuclear spent fuel from people and the environment, protecting them from the heat and radioactivity for a long period of time. After elaboration of a database containing geological, hydrological, tectonic, weather, transport, conservation unit, amongst other information, one intends to combine these information, and make comparisons using preset criteria, in order to indicate the most adequate sites for disposal. (author)

  7. Corrosion resistance of a copper canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    The report presents an evaluation of copper as canister material for spent nuclear fuel. The evaluation is made from the viewpoint of corrosion and applies to a concept of 1977. Supplementary corrosion studies have been performed. The report includes 9 appendices which deal with experimental data. (G.B.)

  8. Studies and research concerning BNFP. Nuclear spent fuel transportation studies

    International Nuclear Information System (INIS)

    Currently, there are a number of institutional problems associated with the shipment of spent fuel assemblies from commercial nuclear power plants: new and conflicting regulations, embargoing of certain routes, imposition of transport safeguards, physical security in-transit, and a lack of definition of when and where the fuel will be moved. This report presents a summary of these types and kinds of problems. It represents the results of evaluations performed relative to fuel receipt at the Barnwell Nuclear Fuel Plant. Case studies were made which address existing reactor sites with near-term spent fuel transportation needs. Shipment by either highway, rail, water, or intermodal water-rail was considered. The report identifies the impact of new regulations and uncertainty caused by indeterminate regulatory policy and lack of action on spent fuel acceptance and storage. This stagnant situation has made it impossible for industry to determine realistic transportation scenarios for business planning and financial risk analysis. A current lack of private investment in nuclear transportation equipment is expected to further prolong the problems associated with nuclear spent fuel and waste disposition. These problems are expected to intensify in the 1980's and in certain cases will make continuing reactor plant operation difficult or impossible

  9. Characterization of alloy particles extracted from spent nuclear fuel

    Science.gov (United States)

    Cui, D.; Rondinella, V. V.; Fortner, J. A.; Kropf, A. J.; Eriksson, L.; Wronkiewicz, D. J.; Spahiu, K.

    2012-01-01

    We characterized, for the first time, submicro- and nanosized fission product-alloy particles that were extracted nondestructively from spent nuclear fuel, in terms of noble metal (Mo-Ru-Tc-Rh-Pd-Te) composition, atomic level homogeneity and lattice parameters. The evidences obtained in this work contribute to an improved understanding of the redox chemistry of radionuclides in nuclear waste repository environments and, in particular, of the catalytic properties of these unique metal alloy particles.

  10. Chemical reprocessing of spent nuclear fuels

    International Nuclear Information System (INIS)

    The reprocessing of nuclear fuels from atomic power stations has a twofold goal. On the one hand it is serving for fuel supply by recovering the fissile materials which have not been consumed or which have been freshly generated in the reactor. On the other hand the radioactive waste products from nuclear power generation are pretreated for long-term safe disposal. The core element of the chemical processing is the PUREX Process, a counter-current solvent extraction procedure using tributyl phosphate (TBP) as the solvent for uranium and plutonium. The chemical basis and the technological performance of the process are discussed. (orig.)

  11. DEVELOPMENT OF METHODOLOGY AND FIELD DEPLOYABLE SAMPLING TOOLS FOR SPENT NUCLEAR FUEL INTERROGATION IN LIQUID STORAGE

    Energy Technology Data Exchange (ETDEWEB)

    Berry, T.; Milliken, C.; Martinez-Rodriguez, M.; Hathcock, D.; Heitkamp, M.

    2012-06-04

    This project developed methodology and field deployable tools (test kits) to analyze the chemical and microbiological condition of the fuel storage medium and determine the oxide thickness on the spent fuel basin materials. The overall objective of this project was to determine the amount of time fuel has spent in a storage basin to determine if the operation of the reactor and storage basin is consistent with safeguard declarations or expectations. This project developed and validated forensic tools that can be used to predict the age and condition of spent nuclear fuels stored in liquid basins based on key physical, chemical and microbiological basin characteristics. Key parameters were identified based on a literature review, the parameters were used to design test cells for corrosion analyses, tools were purchased to analyze the key parameters, and these were used to characterize an active spent fuel basin, the Savannah River Site (SRS) L-Area basin. The key parameters identified in the literature review included chloride concentration, conductivity, and total organic carbon level. Focus was also placed on aluminum based cladding because of their application to weapons production. The literature review was helpful in identifying important parameters, but relationships between these parameters and corrosion rates were not available. Bench scale test systems were designed, operated, harvested, and analyzed to determine corrosion relationships between water parameters and water conditions, chemistry and microbiological conditions. The data from the bench scale system indicated that corrosion rates were dependent on total organic carbon levels and chloride concentrations. The highest corrosion rates were observed in test cells amended with sediment, a large microbial inoculum and an organic carbon source. A complete characterization test kit was field tested to characterize the SRS L-Area spent fuel basin. The sampling kit consisted of a TOC analyzer, a YSI

  12. Criticality safety analysis of WWER-1000 spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Nuclear safety of spent nuclear fuel management is ensured by implementation of the basic safety functions: providing subcriticality, residual heat removal and retention of radioactive products within the physical barriers. To ensure subcriticality during both normal operation and design basis accidents the effective multiplication factor of neutrons Keff must be lower than 0.95. An evaluation of criticality of spent fuel facilities have been made by the modular code system SCALE. The basic calculations are performed with version 6.1 and are validated with version 6.0 of the code system. Spent fuel assemblies type TVSA are modeled as they are representative for WWER-1000 nuclear fuel and cover the characteristics of the earlier modifications of the fuel assemblies. The modeling of the spent fuel containers and equipment is in accordance with actual geometric dimensions and material composition. In all performed calculations, the results demonstrate that the criticality safety criteria are achieved and the effective multiplication factor Keff is lower than the regulatory requirements. (authors)

  13. International tracking and monitoring of nuclear spent fuel transport

    International Nuclear Information System (INIS)

    A cooperative monitoring project was recently initiated to track and monitor shipments of spent fuel from the Lucas Heights HiFAR Research Reactor, located near Sydney, Australia, to either the United States or France. Partners in the project are the US Department of Energy and the Australian Safeguards and Non-Proliferation Office. This project could satisfy a need to have near-real-time continuity of knowledge of a shipment of nuclear material. The benefits of demonstrating a fully operational and sustainable system are significant, because the envisioned system could significantly enhance global safety and security of nuclear material in the transport phases of the nuclear fuel cycle. (author)

  14. OVERVIEW OF CRITERIA FOR INTERIM WET & DRY STORAGE OF RESEARCH REACTOR SPENT NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.; Vinson, D.; Iyer, N.; Fisher, D.

    2010-11-03

    Following discharge from research reactors, spent nuclear fuel may be stored 'wet' in water pools or basins, or it may be stored 'dry' in various configurations including non-sealed or sealed containers until retrieved for ultimate disposition. Interim safe storage practices are based on avoiding degradation to the fuel that would impact functions related to safety. Recommended practices including environmental controls with technical bases, are outlined for wet storage and dry storage of aluminum-clad, aluminum-based research reactor fuel. For wet storage, water quality must be maintained to minimize corrosion degradation of aluminum fuel. For dry storage, vented canister storage of aluminum fuel readily provides a safe storage configuration. For sealed dry storage, drying must be performed so as to minimize water that would cause additional corrosion and hydrogen generation. Consideration must also be given to the potential for radiolytically-generated hydrogen from the bound water in the attendant oxyhydroxides on aluminum fuel from reactor operation for dry storage systems.

  15. Dry Refabrication Technology Development of Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Key technical data on advanced nuclear fuel cycle technology development for the spent fuel recycling have been produced in this study. In the frame work of DUPIC, dry process oxide products fabrication, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remote modulated welding equipment has been designed and fabricated. In the area of advanced pre-treatment process development, a rotary-type oxidizer and spherical particle fabrication process were developed by using SIMFUEL and off-gas treatment technology and zircalloy tube treatment technology were studied. In the area of the property characteristics of dry process products, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data

  16. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, Rodney C.

    2003-09-14

    The successful disposal of spent nuclear fuel (SNF) is one of the most serious challenges to the successful completion of the nuclear fuel cycle and the future of nuclear power generation. In the United States, 21 percent of the electricity is generated by 107 commercial nuclear power plants (NPP), each of which generates 20 metric tons of spent nuclear fuel annually. In 1996, the total accumulation of spent nuclear fuel was 33,700 metric tons of heavy metal (MTHM) stored at 70 sites around the country. The end-of-life projection for current nuclear power plants (NPP) is approximately 86,000 MTHM. In the proposed nuclear waste repository at Yucca Mountain over 95% of the radioactivity originates from spent nuclear fuel. World-wide in 1998, approximately 130,000 MTHM of SNF have accumulated, most of it located at 236 NPP in 36 countries. Annual production of SNF is approximately 10,000 MTHM, containing about 100 tons of ''reactor grade'' plutonium. Any reasonable increase in the proportion of energy production by NPP, i.e., as a substitute for hydrocarbon-based sources of energy, will significantly increase spent nuclear fuel production. Spent nuclear fuel is essentially UO{sub 2} with approximately 4-5 atomic percent actinides and fission product elements. A number of these elements have long half-lives hence, the long-term behavior of the UO{sub 2} is an essential concern in the evaluation of the safety and risk of a repository for spent nuclear fuel. One of the unique and scientifically most difficult aspects of the successful disposal of spent nuclear fuel is the extrapolation of short-term laboratory data (hours to years) to the long time periods (10{sup 3} to 10{sup 5} years) as required by the performance objectives set in regulations, i.e. 10 CFR 60. The direct verification of these extrapolations or interpolations is not possible, but methods must be developed to demonstrate compliance with government regulations and to satisfy the

  17. Review of Drying Methods for Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Large, W.S.

    1999-10-21

    SRTC is developing technology for direct disposal of aluminum spent nuclear fuel (SNF). The development program includes analyses and tests to support design and safe operation of a facility for ''road ready'' dry storage of SNF-filled canisters. The current technology development plan includes review of available SNF drying methods and recommendation of a drying method for aluminum SNF.

  18. Review of partitioning proposals for spent nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bowersox, D.F.

    1976-07-01

    The initial phase of a study about recovery of valuable fission products from spent nuclear fuels has been to review various partitioning proposals. This report briefly describes the aqueous Purex process, the salt transport process, melt refining, fluoride volatility process, and gravimetric separations. All these processes appear to be possible technically, but further research will be necessary to determine which are most feasible. This review includes general recommendations for experimental research and development of several partitioning options.

  19. Spent nuclear fuel canister storage building conceptual design report

    Energy Technology Data Exchange (ETDEWEB)

    Swenson, C.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1996-01-01

    This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ``Technical Baseline and Updated Cost Estimate for the Canister Storage Building``, dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995.

  20. Nonlinear flow effects on immersed spent nuclear racks

    OpenAIRE

    Moreira, Miguel; Antunes, J

    2004-01-01

    Fluid-coupling effects lead to a complex dynamical behavior of immersed spent fuel assembly storage racks. Predicting their responses under strong earthquakes is of prime importance for the safety of nuclear plant facilities. In the near-past we introduced a simplified linearized model for the vibrations of such systems, in which gap-averaged velocity and pressure fields were described analytically in terms of a single space-coordinate for each fluid inter-rack channel. Using such approach it...

  1. Dry store for spent fuel elements from nuclear reactors

    International Nuclear Information System (INIS)

    In the dry store for spent fuel elements from nuclear reactors which are enclosed in storage tubes and cooled with air, the storage tubes being arranged in shafts of a storage building, a loading device is provided underneath the shafts and in a cooling air shaft designed for transporting. The loading device therefore requires only a small lifting height and the chances of storage tubes falling from great heights are excluded. This invention is applicable in particular for intermediate stores. (orig./RW)

  2. Spent nuclear fuel canister storage building conceptual design report

    International Nuclear Information System (INIS)

    This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ''Technical Baseline and Updated Cost Estimate for the Canister Storage Building'', dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995

  3. Naval Spent Nuclear Fuel disposal Container System Description Document

    Energy Technology Data Exchange (ETDEWEB)

    N. E. Pettit

    2001-07-13

    The Naval Spent Nuclear Fuel Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers/waste packages are loaded and sealed in the surface waste handling facilities, transferred underground through the access drifts using a rail mounted transporter, and emplaced in emplacement drifts. The Naval Spent Nuclear Fuel Disposal Container System provides long term confinement of the naval spent nuclear fuel (SNF) placed within the disposal containers, and withstands the loading, transfer, emplacement, and retrieval operations. The Naval Spent Nuclear Fuel Disposal Container System provides containment of waste for a designated period of time and limits radionuclide release thereafter. The waste package maintains the waste in a designated configuration, withstands maximum credible handling and rockfall loads, limits the waste form temperature after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Each naval SNF disposal container will hold a single naval SNF canister. There will be approximately 300 naval SNF canisters, composed of long and short canisters. The disposal container will include outer and inner cylinder walls and lids. An exterior label will provide a means by which to identify a disposal container and its contents. Different materials will be selected for the waste package inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and the natural barrier will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel while the outer cylinder and outer cylinder lids will be made of high-nickel alloy.

  4. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    International Nuclear Information System (INIS)

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  5. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  6. Standard guide for drying behavior of spent nuclear fuel

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This guide is organized to discuss the three major components of significance in the drying behavior of spent nuclear fuel: evaluating the need for drying, drying spent nuclear fuel, and confirmation of adequate dryness. 1.1.1 The guide addresses drying methods and their limitations in drying spent nuclear fuels that have been in storage at water pools. The guide discusses sources and forms of water that remain in SNF, its container, or both, after the drying process and discusses the importance and potential effects they may have on fuel integrity, and container materials. The effects of residual water are discussed mechanistically as a function of the container thermal and radiological environment to provide guidance on situations that may require extraordinary drying methods, specialized handling, or other treatments. 1.1.2 The basic issue in drying is to determine how dry the SNF must be in order to prevent issues with fuel retrievability, container pressurization, or container corrosion. Adequate d...

  7. Studies on spent nuclear fuel evolution during storage

    Energy Technology Data Exchange (ETDEWEB)

    Rondinella, V.V.; Wiss, T.A.G.; Papaioannou, D.; Nasyrow, R. [European Commission Joint Research Centre, Karlsruhe (Germany). Inst. for Transuranium Elements

    2015-07-01

    Initially conceived to last only a few decades (40 years in Germany), extended storage periods have now to be considered for spent nuclear fuel due to the expanding timeline for the definition and implementation of the disposal in geologic repository. In some countries, extended storage may encompass a timeframe of the order of centuries. The safety assessment of extended storage requires predicting the behavior of the spent fuel assemblies and the package systems over a correspondingly long timescale, to ensure that the mechanical integrity and the required level of functionality of all components of the containment system are retained. Since no measurement of ''old'' fuel can cover the ageing time of interest, spent fuel characterization must be complemented by studies targeting specific mechanisms that may affect properties and behavior of spent fuel during extended storage. Tests conducted under accelerated ageing conditions and other relevant simulations are useful for this purpose. During storage, radioactive decay determines the overall conditions of spent fuel and generates heat that must be dissipated. Alpha-decay damage and helium accumulation are key processes affecting the evolution of properties and behavior of spent fuel. The radiation damage induced by a decay event during storage is significantly lower than that caused by a fission during in-pile operation: however, the duration of the storage is much longer and the temperature levels are different. Another factor potentially affecting the mechanical integrity of spent fuel rods during storage and handling / transportation is the behavior of hydrogen present in the cladding. At the Institute for Transuranium Elements, part of the Joint Research Centre of the European Commission, spent fuel alterations as a function of time and activity are monitored at different scales, from the microstructural level (defects and lattice parameter swelling) up to macroscopic properties such as

  8. Report on interim storage of spent nuclear fuel. Midwestern high-level radioactive waste transportation project

    Energy Technology Data Exchange (ETDEWEB)

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  9. 77 FR 37937 - License Renewal Application for Prairie Island Nuclear Generating Plant Independent Spent Fuel...

    Science.gov (United States)

    2012-06-25

    ..., Division of Spent Fuel Storage and Transportation, Office of Nuclear Material Safety and Safeguards, U.S..., possession, storage and transfer of spent fuel, reactor-related Greater than Class C (GTCC) waste and other radioactive materials associated with spent fuel storage at the PINGP site-specific Independent Spent......

  10. DEMONSTRATION OF LONG-TERM STORAGE CAPABILITY FOR SPENT NUCLEAR FUEL IN L BASIN

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.; Deible, R.

    2011-04-27

    The U.S. Department of Energy decisions for the ultimate disposition of its inventory of used nuclear fuel presently in, and to be received and stored in, the L Basin at the Savannah River Site, and schedule for project execution have not been established. A logical decision timeframe for the DOE is following the review of the overall options for fuel management and disposition by the Blue Ribbon Commission on America's Nuclear Future (BRC). The focus of the BRC review is commercial fuel; however, the BRC has included the DOE fuel inventory in their review. Even though the final report by the BRC to the U.S. Department of Energy is expected in January 2012, no timetable has been established for decisions by the U.S. Department of Energy on alternatives selection. Furthermore, with the imminent lay-up and potential closure of H-canyon, no ready path for fuel disposition would be available, and new technologies and/or facilities would need to be established. The fuel inventory in wet storage in the 3.375 million gallon L Basin is primarily aluminum-clad, aluminum-based fuel of the Materials Test Reactor equivalent design. An inventory of non-aluminum-clad fuel of various designs is also stored in L Basin. Safe storage of fuel in wet storage mandates several high-level 'safety functions' that would be provided by the Structures, Systems, and Components (SSCs) of the storage system. A large inventory of aluminum-clad, aluminum-based spent nuclear fuel, and other nonaluminum fuel owned by the U.S. Department of Energy is in wet storage in L Basin at the Savannah River Site. An evaluation of the present condition of the fuel, and the Structures, Systems, or Components (SSCs) necessary for its wet storage, and the present programs and storage practices for fuel management have been performed. Activities necessary to validate the technical bases for, and verify the condition of the fuel and the SSCs under long-term wet storage have also been identified. The

  11. 75 FR 60147 - Calvert Cliffs Nuclear Power Plant, LLC; Independent Spent Fuel Storage Installation; Notice of...

    Science.gov (United States)

    2010-09-29

    ... COMMISSION Calvert Cliffs Nuclear Power Plant, LLC; Independent Spent Fuel Storage Installation; Notice of... Branch, Division of Spent Fuel Storage and Transportation, Office of Nuclear Material Safety and... Independent Spent Fuel Storage Installation (ISFSI) Technical Specifications (TS) be revised as follows: 1....

  12. Spent nuclear fuel for disposal in the KBS-3 repository

    Energy Technology Data Exchange (ETDEWEB)

    Grahn, Per; Moren, Lena; Wiborgh, Maria

    2010-12-15

    The report is included in a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The set of reports is included in the safety report for the KBS-3 repository and repository facility. The report provides input to the assessment of the long-term safety, SR-Site as well as to the operational safety report, SR-Operation. The report presents the spent fuel to be deposited, and the requirements on the handling and selection of fuel assemblies for encapsulation that follows from that it shall be deposited in the KBS-3 repository. An overview of the handling and a simulation of the encapsulation and the resulting canisters to be deposited are presented. Finally, the initial state of the encapsulated spent nuclear fuel is given. The initial state comprises the radionuclide inventory and other data required for the assessment of the long-term safety

  13. Cost estimations for deep disposal of spent nuclear fuels

    International Nuclear Information System (INIS)

    According to the Act on the Financing of Future Expenses for Spent Nuclear Fuel etc. (Financing Act), the Swedish Nuclear Fuel and Waste Management Co. (SKB) must submit, every year, to the Swedish Nuclear Power Inspectorate (SKI), a cost estimate for the management of spent nuclear fuel and for the decommissioning and dismantling of the nuclear power plants. After SKI has examined and evaluated the cost estimates, SKI must submit a proposal to the Government concerning the fee which should be paid by the nuclear power companies per kWh of generated electricity. According to the Financing Act, the reactor owners must pledge collateral in the event that the accumulated fees should be found to be insufficient as a result of early closure of reactors or as a result of underestimating the future expenses of managing the spent nuclear fuel and of decommissioning and dismantling the reactors. The future total expenses resulting from the Financing Act are estimated at about SEK 48 billion at the January 1998 price level. Of this amount, the cost of the final disposal of spent nuclear fuel in SKB's programme is expected to amount to about SEK 12 billion. SKB's estimate comprises the cost of siting, construction and operation of a deep repository for spent nuclear fuel, based on the KBS-3 concept, and a rock cavern for other long-lived waste which SKB plans to locate next to the spent fuel repository. The cost estimate also includes the dismantling and closure of the facility once all of the fuel and the long-lived waste are deposited. The calculations are based on all of the fuel, which will be generated through the operation of the 12 Swedish reactors during a period of 25 years and for every additional year of operation. At the beginning of 1998, SKI commissioned BERGAB to evaluate the cost estimate for the deep disposal of the spent nuclear fuel. The task was divided into two stages, namely a study which was submitted in June 1998 concerning the technical feasibility of

  14. Spent nuclear fuel management. Moving toward a century of spent fuel management: A view from the halfway mark

    International Nuclear Information System (INIS)

    Full text: A half-century ago, President Eisenhower in his 1953 'Atoms for Peace' speech, offered nuclear technology to other nations as part of a broad nuclear arms control initiative. In the years that followed, the nuclear power generation capabilities of many nations has helped economic development and contributed to the prosperity of the modern world. The growth of nuclear power, while providing many benefits, has also contributed to an increasing global challenge over safe and secure spent fuel management. Over 40 countries have invested in nuclear energy, developing over 400 nuclear power reactors. Nuclear power supplies approximately 16% of the global electricity needs. With the finite resources and challenges of fossil fuels, nuclear power will undoubtedly become more prevalent in the future, both in the U.S. and abroad. We must address this inevitability with new paradigms for managing a global nuclear future. Over the past fifty years, the world has come to better understand the strong interplay between all elements of the nuclear fuel cycle, global economics, and global security. In the modern world, the nuclear fuel cycle can no longer be managed as a simple sequence of technological, economic and political challenges. Rather it must be seen, and managed, as a system of strongly interrelated challenges. Spent fuel management, as one element of the nuclear fuel system, cannot be relegated to the back-end of the fuel cycle as only a disposal or storage issue. There exists a wealth of success and experience with spent fuel management over the past fifty years. We must forge this experience with a global systems perspective, to reshape the governing of all aspects of the nuclear fuel cycle, including spent fuel management. This session will examine the collective experience of spent fuel management enterprises, seeking to shape the development of new management paradigms for the next fifty years. (author)

  15. Development of dry storage technology of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Maruoka, Kunio [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Nuclear Energy Systems Engineering Center; Murakami, Kazuo; Yokoyama, Takeshi; Natsume, Tomohiro; Irino, Mitsuhiro

    1998-07-01

    The increasing demand for storage of spent fuel assemblies generated by commercial nuclear power plants is the urgent subject to solve. The dry storage system is as economically more advantageous than the pool storage system, and so, Mitsubishi Heavy Industries, Ltd. has developed the metal storage cask suited to small and medium storage capacity under 2000MTU - 3000MTU. For large scale capacity, the new `Mitsubishi Vault Storage System` has been developed, and it provides a safe and economical solution. Technical study concerning cooling ability was performed. (author)

  16. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    International Nuclear Information System (INIS)

    Spent nuclear fuel, essentially U2, accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO2 in SNF is not stable under oxiding conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term

  17. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Rodney C. Ewing

    2004-10-07

    Spent nuclear fuel, essentially U{sub 2}, accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO{sub 2} in SNF is not stable under oxiding conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term.

  18. Dosimetry at an interim storage for spent nuclear fuel.

    Science.gov (United States)

    Králík, M; Kulich, V; Studeny, J; Pokorny, P

    2007-01-01

    The Czech nuclear power plant Dukovany started its operation in 1985. All fuel spent from 1985 up to the end of 2005 is stored at a dry interim storage, which was designed for 60 CASTOR-440/84 casks. Each of these casks can accommodate 84 fuel assemblies from VVER 440 reactors. Neutron-photon mixed fields around the casks were characterized in terms of ambient dose equivalent measured by standard area dosemeters. Except this, neutron spectra were measured by means of a Bonner sphere spectrometer, and the measured spectra were used to derive the corresponding ambient dose equivalent due to neutrons. PMID:17526479

  19. Dosimetry at an interim storage for spent nuclear fuel.

    Science.gov (United States)

    Králík, M; Kulich, V; Studeny, J; Pokorny, P

    2007-01-01

    The Czech nuclear power plant Dukovany started its operation in 1985. All fuel spent from 1985 up to the end of 2005 is stored at a dry interim storage, which was designed for 60 CASTOR-440/84 casks. Each of these casks can accommodate 84 fuel assemblies from VVER 440 reactors. Neutron-photon mixed fields around the casks were characterized in terms of ambient dose equivalent measured by standard area dosemeters. Except this, neutron spectra were measured by means of a Bonner sphere spectrometer, and the measured spectra were used to derive the corresponding ambient dose equivalent due to neutrons.

  20. TVO's new encapsulation method for spent nuclear fuel

    International Nuclear Information System (INIS)

    Teollisuuden Voima Oy has developed a new encapsulation method for spent nuclear fuel s.c. cold process. Instead of casting (400 deg C) molten lead the new canister is filled with cold granulated material like quartz sand, lead shots or glassa beads. The new canister concept ACPC (Advanced Cold Process Canister) consists of the outer oxygen free copper canister and of the inner steel canister. The function of the steel canister is merely to give mechanical strength. The copper canister acts as corrosion protection guaranteeing practically lifetime of millions of years for the ACPC concept

  1. Challenges in spent nuclear fuel final disposal:conceptual design models

    Institute of Scientific and Technical Information of China (English)

    Mukhtar Ahmed RANA

    2008-01-01

    The disposal of spent nuclear fuel is a long-standing issue in nuclear technology. Mainly, UO2 and metallic U are used as a fuel in nuclear reactors. Spent nuclear fuel contains fission products and transuranium elements, which would remain radioactive for 104 to 108 years. In this brief communication, essential concepts and engineering elements related to high-level nuclear waste disposal are described. Conceptual design models are described and discussed considering the long-time scale activity of spent nuclear fuel or high level waste. Notions of physical and chemical barriers to contain nuclear waste are highlightened. Concerns regarding integrity, self-irradiation induced decomposition and thermal effects of decay heat on the spent nuclear fuel are also discussed. The question of retrievability of spent nuclear fuel after disposal is considered.

  2. Spent nuclear fuel policies in historical perspective: An international comparison

    International Nuclear Information System (INIS)

    The purpose of this article is to explain why the world's nuclear power countries differ from each other with respect to their spent nuclear fuel (SNF) policies. The emergence and evolution of three principal SNF approaches are analyzed: direct disposal, reprocessing and SNF export. Five broad explanatory factors are identified and discussed in relation to the observed differences in policy outcomes: military ambitions and non-proliferation, technological culture, political culture and civil society, geological conditions, and energy policy. SNF policy outcomes can generally be seen to result from a complex interaction between these broad factors, but it is also possible to discern a number of important patterns. To the extent that the five factors may undergo far-reaching changes in the future, the historical experience of how they have shaped SNF policies also give a hint of possible future directions in SNF policymaking around the world

  3. Training implementation matrix, Spent Nuclear Fuel Project (SNFP)

    Energy Technology Data Exchange (ETDEWEB)

    EATON, G.L.

    2000-06-08

    This Training Implementation Matrix (TIM) describes how the Spent Nuclear Fuel Project (SNFP) implements the requirements of DOE Order 5480.20A, Personnel Selection, Qualification, and Training Requirements for Reactor and Non-Reactor Nuclear Facilities. The TIM defines the application of the selection, qualification, and training requirements in DOE Order 5480.20A at the SNFP. The TIM also describes the organization, planning, and administration of the SNFP training and qualification program(s) for which DOE Order 5480.20A applies. Also included is suitable justification for exceptions taken to any requirements contained in DOE Order 5480.20A. The goal of the SNFP training and qualification program is to ensure employees are capable of performing their jobs safely and efficiently.

  4. Spent nuclear fuel recycling with plasma reduction and etching

    Science.gov (United States)

    Kim, Yong Ho

    2012-06-05

    A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF6 and PuF4. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF6 (gas) that is formed to decompose back to PuF4 (solid), and b) to maintain stability of the UF6. Uranium (in the form of gaseous UF6) is easily extracted and separated from the plutonium (in the form of solid PuF4). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF6 (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF4 or NF3 as the fluorine sources instead of F2 or HF.

  5. Physical modeling of spent-nuclear-fuel container

    Directory of Open Access Journals (Sweden)

    Wang Liping

    2012-11-01

    Full Text Available A new physical simulation model was developed to simulate the casting process of the ductile iron heavy section spent-nuclear-fuel container. In this physical simulation model, a heating unit with DR24 Fe-Cr-Al heating wires was used to compensate the heat loss across the non-natural surfaces of the sample, and a precise and reliable casting temperature controlling/monitoring system was employed to ensure the thermal behavior of the simulated casting to be similar to the actual casting. Also, a mould system was designed, in which changeable mould materials can be used for both the outside and inside moulds for different applications. The casting test was carried out with the designed mould and the cooling curves of central and edge points at different isothermal planes of the casting were obtained. Results show that for most isothermal planes, the temperature control system can keep the temperature differences within 6 ℃ between the edge points and the corresponding center points, indicating that this new physical simulation model has high simulation accuracy, and the mould developed can be used for optimization of casting parameters of spent-nuclear-fuel container, such as composition of ductile iron, the pouring temperature, the selection of mould material and design of cooling system. In addition, to maintain the spheroidalization of the ductile iron, the force-chilling should be used for the current physical simulation to ensure the solidification of casting in less than 2 h.

  6. Cost and risk tradeoff for routing nuclear spent fuel movements

    International Nuclear Information System (INIS)

    In the transportation industry, much effort has been devoted to finding the least cost routes for shipping goods from their production sites to the market areas. In addition to cost, the decision maker must take the risk of an incident into consideration for transportation routing involving hazardous materials. The transportation of spent nuclear fuel from reactor sites to repositories is an example. Given suitable network information, existing routing methods can readily determine least cost or least risk routes for any shipment. These two solutions, however, represent the extremes of a large number of alternatives with different combinations of risk and cost. In the selection of routes and also in the evaluation of alternative storage sites it is not enough to know which is the lease cost or lowest risk. Intelligent decision-marking requires knowledge of how much it will cost to lower risk by a certain amount. The objective of this study is to develop an automated system to evaluate the tradeoff between transportation cost and potential population at risk under different nuclear spent fuel transportation strategies

  7. K-Basin spent nuclear fuel characterization data report 2

    International Nuclear Information System (INIS)

    An Integrated Process Strategy has been developed to package, condition, transport, and store in an interim storage facility the spent nuclear fuel (SNF) currently residing in the K-Basins at Hanford. Information required to support the development of the condition process and to support the safety analyses must be obtained from characterization testing activities conducted on fuel samples from the Basins. Some of the information obtained in the testing was reported in PNL-10778, K-Basin Spent Nuclear Fuel Characterization Data Report (Abrefah et al. 1995). That report focused on the physical, dimensional, metallographic examinations of the first K-West (KW) Basin SNF element to be examined in the Postirradiation Testing Laboratory (PTL) hot cells; it also described some of the initial SNF conditioning tests. This second of the series of data reports covers the subsequent series of SNF tests on the first fuel element. These tests included optical microscopy analyses, conditioning (drying and oxidation) tests, ignition tests, and hydrogen content tests

  8. Thermal hydraulic feasibility assessment for the Spent Nuclear Fuel Project

    International Nuclear Information System (INIS)

    A series of scoping analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The SNFP was established to develop engineered solutions for the expedited removal, stabilization, and storage of spent nuclear fuel from the K Basins at the U.S. Department of Energy's Hanford Site in Richland, Washington. The subject efforts focused on independently investigating, quantifying, and establishing the governing heat production and removal mechanisms for each of the IPS operations and configurations, obtaining preliminary results for comparison with and verification of other analyses, and providing technology-based recommendations for consideration and incorporation into the design bases for the SNFP. The goal was to develop a series fo thermal-hydraulic models that could respond to all process and safety-related issues that may arise pertaining to the SNFP. A series of sensitivity analyses were also performed to help identify those parameters that have the greatest impact on energy transfer and hence, temperature control. It is anticipated that the subject thermal-hydraulic models will form the basis for a series of advanced and more detailed models that will more accurately reflect the thermal performance of the IPS and alleviate the necessity for some of the more conservative assumptions and oversimplifications, as well as form the basis for the final process and safety analyses

  9. 78 FR 20625 - Spent Nuclear Fuel Management at the Savannah River Site

    Science.gov (United States)

    2013-04-05

    ... of Decision (ROD) pursuant to the Savannah River Site Spent Nuclear Fuel Management Final... Concerning Foreign Research Reactor Spent Nuclear Fuel Environmental Impact Statement (DOE/EIS-0218, 1996... to create LEU feedstock for fuel fabrication for commercial nuclear reactors. The shipments of...

  10. Impact evaluation of sea salt aerosol given to nuclear spent fuel intermediate storage facility

    International Nuclear Information System (INIS)

    Nuclear spent fuel intermediate storage facility is planned to be located in the Sea of Japan coast area for the reasons of nuclear spent fuel transportation. Therefore, there is the risk of structure corrosion by an environmental pollution factor e.g. sea salt aerosol. This study is the corrosion risk evaluation of nuclear spent fuel storage facility by measuring the amount of sea salt aerosol coming. This study develops the new method to improve of the conventional JIS method. (author)

  11. Disposal of spent fuel from German nuclear power plants - 16028

    International Nuclear Information System (INIS)

    The 'direct disposal of spent fuel' as a part of the current German reference concept was developed as an alternative to spent fuel reprocessing and vitrified HLW disposal. The technical facilities necessary for the implementation of this part of the reference concept, the so called POLLUXR concept, i.e. interim storage buildings for casks containing spent fuel, a pilot conditioning facility, and a special cask 'POLLUX' for final disposal have been built. With view to a geological salt formation all handling procedures for the direct disposal of spent fuel were tested aboveground in full-scale test facilities. To optimise the reference concept, all operational steps have been reviewed for possible improvements. The two additional concepts for the direct disposal of SF are the BSK 3 concept and the DIREGT concept. Both concepts rely on borehole emplacement technology, vertical boreholes for the BSK 3 concept und horizontal boreholes for the DIREGT concept. Supported by the EU and the German Federal Ministry of Economics and Technology (BMWi), DBE TECHNOLOGY built an aboveground full-scale test facility to simulate all relevant handling procedures for the BSK 3 disposal concept. GNS (Company for Nuclear Service), representing the German utilities, provided the main components and its know-how concerning cask design and manufacturing. The test program was concluded recently after more than 1.000 emplacement operations had been performed successfully. The BSK 3 emplacement system in total comprises an emplacement device, a borehole lock, a transport cart, a transfer cask which will shuttle between the aboveground conditioning facility and the underground repository, and the BSK 3 canister itself, designed to contain the fuel rods of three PWR-fuel assemblies with a total of about 1.6 tHM. The BSK 3 concept simplifies the operation of the repository because the handling procedures and techniques can also be applied for the disposal of reprocessing residues. In addition

  12. 78 FR 61401 - Entergy Nuclear Operations, Inc.; Big Rock Point; Independent Spent Fuel Storage Installation

    Science.gov (United States)

    2013-10-03

    ... COMMISSION Entergy Nuclear Operations, Inc.; Big Rock Point; Independent Spent Fuel Storage Installation..., Inc. (ENO) on June 20, 2012, for the Big Rock Point (BRP) Independent Spent Fuel Storage Installation... Regulatory Evaluation In the Final Rule for Storage of Spent Fuel in NRC-Approved Storage Casks at...

  13. 76 FR 22935 - Calvert Cliffs Nuclear Power Plant, LLC Independent Spent Fuel Storage Installation; Notice of...

    Science.gov (United States)

    2011-04-25

    ... Class C (GTCC) waste and other radioactive materials associated with spent fuel storage at the CCNPP... granted, the renewed license will authorize the applicant to continue to store spent fuel in a dry cask... COMMISSION Calvert Cliffs Nuclear Power Plant, LLC Independent Spent Fuel Storage Installation; Notice...

  14. System evaluation model for selecting spent nuclear fuel storage concepts

    International Nuclear Information System (INIS)

    President Reagan has challenged the nuclear community to develop a safe and economical system to handle spent fuel assemblies from commercial nuclear power generating stations and high level radioactive wastes from fuel reprocessing facilities. This paper will illustrate a management system approach used to identify and evaluate monitored, retrievable fuel storage concepts that fulfill ten key criteria for meeting the functional requirements and system objectives of the National Nuclear Waste Management Program. These selection criteria include: health and safety, schedule, costs, socio-economic factors and environmental factors. An MRS (Monitored Retrievable Storage) System is designed to safety store spent fuel assemblies and high level waste in a dry, passively cooled environment for an indefinite period of time up to 100 years. Forth-five feasible MRS design concepts utilizing 20 basic storage methods were initially screened and reduced to 15 concepts with 7 methods. A system evaluation model was developed to evaluate the ability of each of the remaining 15 alternative concepts to meet the selection criteria. This model allows the ranking of each system according to MRS needs and life-cycle costs. The methodology used to establish the selection criteria, develop a weight of importance for each criterion and assess the relative merit of each MRS system will be described. The impact of cost relative to technical criteria will be discussed along with experience in obtaining relative merit data and its application in the model. This system evaluation model is universally applicable when many concepts in various stages of design and cost development need to be evaluated. It provides a structured approach which allows the evaluators to make a rational, well-founded selection

  15. IAEA nuclear fuel cycle databases: Relevance to spent nuclear fuel management

    International Nuclear Information System (INIS)

    Full text: Reliable statistical data on spent fuel management would be essential for the global nuclear community, e.g. for approaches related to international cooperation, as well as for the needs of individual countries. Compilation of data on large amounts of spent fuel located at various nuclear facilities around the world is a challenge. It is not a trivial exercise to collect and compile spent fuel inventory data as they are subject to dynamic change. Spent fuel inventory data are important to various national and international spent fuel management activities, especially for planning and regulatory activities. Recently, security issues became an additional factor to be considered in the information management associated with spent fuel or radioactive waste. The specific need for spent fuel inventory data varies depending on the ultimate purpose: International Level - compilation on a gross tonnage (in heavy metal basis) mainly for statistical purposes and global trend analysis both for use by IAEA and at the request of Member States; National Level - compilation for industry and regulatory purposes on either a gross tonnage or individual assembly basis to assist in planning and public awareness; and Operator Level - the origination and maintenance of detailed data on individual assemblies by the utility for operational needs or to meet regulatory requirements. There is, in general, a global trend towards greater transparency of information with the general public which may require more information to be made public on spent fuel management, including data on inventories or transportation. With the increase in the commercialisation of the nuclear industry, the trend is away from national governments operating nuclear facilities, including spent fuel management. This results in the spread of information on spent fuel as it is not concentrated at government level, but is instead held by various organizations . Spent fuel information may also have to be

  16. Damage in spent nuclear fuel defined by properties and requirements

    International Nuclear Information System (INIS)

    Full text: The properties of light water reactor (LWR) fuel rods and assemblies are altered in service due to irradiation. Some of these alterations render the fuel unsuitable for emplacement in casks used for storage or transportation without special handling. Title 10 (Energy) of the U.S. Code of Federal Regulations Part 72 (storage) and Part 71 (transportation) establish direct requirements for the behavior expected of spent fuel. In particular, retrievability and prevention of gross breaches are required in storage and no reconfiguration of the fuel is allowed during normal transport. In addition, in the process of meeting other regulations related to criticality, shielding, and containment, the cask designers may need to place additional requirements on the behavior of the fuel. The definition of damaged fuel might be based on the ability of the fuel to perform in a manner such that the direct regulatory requirements and the onus placed on the fuel by the cask designer are met. Fuels that have alterations that do not permit it to perform its required safety function, without special handling, should be regarded as damaged. Since the requirements placed on the fuel may vary during phases of the fuel cycle, the potential exists for independent definitions to co-exist for interim dry storage, transport, and final disposal in a geologic repository. The United States Nuclear Regulatory Commission's (USNRC) Spent Fuel Program Office (SFPO) has provided guidance in defining damaged fuel in Interim Staff Guidance (ISG) -1. This guidance is similar to that being developed by the American National Standards Institute (ANSI). Neither of these documents provides the logic behind the definition of damaged fuel. This paper will discuss the requirements placed on the fuel for dry interim storage and transportation and the ways that these requirements drive the definition of damaged spent fuel. Examples will be given illustrating the methodology. (author)

  17. Seismic analysis of spent nuclear fuel storage racks

    Energy Technology Data Exchange (ETDEWEB)

    Shah, S.J.; Biddle, J.R.; Bennett, S.M.; Schechter, C.B. [Framatome Cogema Fuels, Lynchburg, VA (United States); Harstead, G.A. [Harstead Engineering Associates, Inc., Old Tappan, NJ (United States); Marquet, F. [ATEA/FRAMATOME, Carquefou (France)

    1996-06-01

    In many nuclear power plants, existing storage racks are being replaced with high-density racks to accommodate the increasing inventory of spent fuel. In the hypothetical design considered here, the high-density arrangement of fuel assemblies, or consolidated fuel canisters, is accomplished through the use of borated stainless steel (BSS) plates acting as neutron absorbers. No structural benefit from the BSS is assumed. This paper describes the methods used to perform seismic analysis of high density spent fuel storage racks. The sensitivity of important parameters such as the effect of variation of coefficients of friction between the rack legs and the pool floor and fuel loading conditions (consolidated and unconsolidated) are also discussed in the paper. Results of this study are presented. The high-density fuel racks are simply supported by the pool floor with no structural connections to adjacent racks or to the pool walls or floor. Therefore, the racks are free standing and may slide and tip. Several time history, nonlinear, seismic analyses are required to account for variations in the coefficient of friction, rack loading configuration, and the type of the seismic event. This paper presents several of the mathematical models usually used. Friction cannot be precisely predicted, so a range of friction coefficients is assumed. The range assumed for the analysis is 0.2 to 0.8. A detailed model representing a single rack is used to evaluate the 3-D loading effects. This model is a controlling case for the stress analysis. A 2-D multi-rack model representing a row of racks between the spent fuel pool walls is used to evaluate the change in gaps between racks. The racks are normally analyzed for the fuel loading conditions of consolidated, full, empty, and half-loaded with fuel assemblies.

  18. DOE-owned spent nuclear fuel strategic plan. Revision 1

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) is responsible for safely and efficiently managing DOE-owned spent nuclear fuel (SNF) and SNF returned to the US from foreign research reactors (FRR). The fuel will be treated where necessary, packaged suitable for repository disposal where practicable, and placed in interim dry storage. These actions will remove remaining vulnerabilities, make as much spent fuel as possible ready for ultimate disposition, and substantially reduce the cost of continued storage. The goal is to complete these actions in 10 years. This SNF Strategic Plan updates the mission, vision, objectives, and strategies for the management of DOE-owned SNF articulated by the SNF Strategic Plan issued in December 1994. The plan describes the remaining issues facing the EM SNF Program, lays out strategies for addressing these issues, and identifies success criteria by which program progress is measured. The objectives and strategies in this plan are consistent with the following Em principles described by the Assistance Secretary in his June 1996 initiative to establish a 10-year time horizon for achieving most program objectives: eliminate and manage the most serious risks; reduce mortgage and support costs to free up funds for further risk reduction; protect worker health and safety; reduce generation of wastes; create a collaborative relationship between DOE and its regulators and stakeholders; focus technology development on cost and risk reduction; and strengthen management and financial control

  19. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Matthew D. Hinds

    2001-10-17

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.

  20. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  1. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array

  2. DOE-owned spent nuclear fuel strategic plan. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    The Department of Energy (DOE) is responsible for safely and efficiently managing DOE-owned spent nuclear fuel (SNF) and SNF returned to the US from foreign research reactors (FRR). The fuel will be treated where necessary, packaged suitable for repository disposal where practicable, and placed in interim dry storage. These actions will remove remaining vulnerabilities, make as much spent fuel as possible ready for ultimate disposition, and substantially reduce the cost of continued storage. The goal is to complete these actions in 10 years. This SNF Strategic Plan updates the mission, vision, objectives, and strategies for the management of DOE-owned SNF articulated by the SNF Strategic Plan issued in December 1994. The plan describes the remaining issues facing the EM SNF Program, lays out strategies for addressing these issues, and identifies success criteria by which program progress is measured. The objectives and strategies in this plan are consistent with the following Em principles described by the Assistance Secretary in his June 1996 initiative to establish a 10-year time horizon for achieving most program objectives: eliminate and manage the most serious risks; reduce mortgage and support costs to free up funds for further risk reduction; protect worker health and safety; reduce generation of wastes; create a collaborative relationship between DOE and its regulators and stakeholders; focus technology development on cost and risk reduction; and strengthen management and financial control.

  3. Patent Analysis for Pyroprocessing of Spent Nuclear Fuels

    International Nuclear Information System (INIS)

    Analysis of foreign and domestic patents for pyroprocessing technology of spent nuclear fuels was carried out in this study. The current status of pyroprocessing technology development in such countries as Korea, USA, Japan and EU was analyzed by classifying the patents for 1975 through 2009 according to registration country, assignee, calendar year and technology area. The major assignees' activity indices were compared in order to find out whether there is any concentrated area of technical details. Technology competitiveness of the countries was also investigated from the information of patent citation number and family size. Furthermore, some essential unit technologies required for the commercialization of pyroprocessing were derived and examined in the aspect of the state of art as well as the trend of technology development.

  4. Spent Nuclear Fuel Project Cold Vacuum Drying Facility Operations Manual

    International Nuclear Information System (INIS)

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-553, Spent Nuclear Fuel Project Final Safety Analysis Report Annex B--Cold Vacuum Drying Facility. The HNF-SD-SNF-DRD-002, 1999, (Cold Vacuum Drying Facility Design Requirements), Rev. 4. and the CVDF Final Design Report. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence and references to the CVDF System Design Descriptions (SDDs). This manual has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  5. Retrievability of spent nuclear fuel canisters; Kaeytetyn ydinpolttoaineen loppusijoituskapseleiden palautettavuus

    Energy Technology Data Exchange (ETDEWEB)

    Saanio, T. [Saanio and Riekkola Oy, Helsinki (Finland); Raiko, H. [VTT Energy, Espoo (Finland)

    1999-03-01

    As a part of the designing process of the Finnish spent nuclear fuel repository, a preliminary study has been carried out to investigate how the canisters could technically be retrieved to the ground surface. Possibility of retrieving a canister has been investigated in different phases of the disposal project. Retrievability has not been a design goal for the spent fuel repository. However, design of the repository includes some features that may ease the retrieval of canisters in the future. Spent fuel elements are packaged in massive copper-iron canisters, which are mechanically strong and long-lived. The repository consists of excavated tunnels in hard rock which are supposed to be very long-lived making the removal of the tunnel backfilling technically possible also in the future. As long as the bentonite buffer has not been installed the canister can be returned to the ground surface using the same equipment as was used when the canister was brought down to the repository and lowered into the hole. In the encapsulation station the spent fuel elements can be packaged in the other canister or in the transport cask. After a deposition tunnel has been backfilled and closed, the retrieval consists of tearing down the concrete structure at the entry of the deposition tunnel, removal of the tunnel backfilling, removal of the bentonite from the disposal hole and lifting up of the canister. Various methods, e.g., flushing the bentonite with saline solutions, can be used to detach the canister from a hole with fully saturated bentonite. Recovery will be technically possible also after closing of the disposal facility. Backfilling of the shafts and tunnels will be removed and additional new structures and systems will have to be built in the repository. After that canisters can be transported to the ground surface as described above. In addition, handling of the canisters at the ground surface will require additional facilities. Canisters can be packaged in the

  6. Storage of Spent Nuclear Fuel in Norway: Status and Prospects

    International Nuclear Information System (INIS)

    Spent Nuclear Fuel (SNF) in Norway has arisen from irradiation of fuel in the JEEP I and JEEP II reactors at Kjeller, and in the Halden Boiling Water Reactor (HBWR) in Halden. In total there are some 16 tonnes of SNF, all of which is currently stored on-site, in either wet or dry storage facilities. The greater part of the SNF, 12 tonnes, consists of aluminium-clad fuel, of which 10 tonnes is metallic uranium fuel and the remainder oxide (UO2). Such fuel presents significant challenges with respect to long-term storage and disposal. Current policy is that existing spent fuel will, as far as possible considering its suitability for later direct disposal, be stored until final disposal is possible. Several committees have advised the Government of Norway on, among others, policy issues, storage methods and localisation of a storage facility. Both experts and stakeholders have participated in these committees. This paper presents an overview of the spent fuel in Norway and a description of current storage arrangements. The prospects for long-term storage are then described, including a summary of recommendations made to government, the reactions of various stakeholders to these recommendations, the current status, and the proposed next steps. A recommended policy is to construct a new storage facility for the fuel to be stored for a period of at least 50 years. In the meantime a national final disposal facility should be constructed and taken into operation. It has been recommended that the aluminium-clad fuel be reprocessed in an overseas commercial facility to produce a stable waste form for storage and disposal. This recommendation is controversial, and a decision has not yet been taken on whether to pursue this option. An analysis of available storage concepts for the more modern fuel types resulted in the recommendation to use dual-purpose casks. In addition, it was recommended to construct a future storage facility in a rock hall instead of a free

  7. Reduction of uranium in disposal conditions of spent nuclear fuel

    International Nuclear Information System (INIS)

    This literature study is a summary of publications, in which the reduction of uranium by iron has been investigated in anaerobic groundwater conditions or in aqueous solution in general. The basics of the reduction phenomena and the oxidation states, complexes and solubilities of uranium and iron in groundwaters are discussed as an introduction to the subject, as well as, the Finnish disposal concept of spent nuclear fuel. The spent fuel itself mainly (∼96 %) consists of a sparingly soluble uranium(IV) dioxide, UO2(s), which is stable phase in the anticipated reducing disposal conditions. If spent fuel gets in contact with groundwater, oxidizing conditions might be induced by the radiolysis of water, or by the intrusion of oxidizing glacial melting water. Under these conditions, the oxidation and dissolution of uranium dioxide to more soluble U(VI) species could occur. This could lead to the mobilization of uranium and other components of spent fuel matrix including fission products and transuranium elements. The reduction of uranium back to oxidation state U(IV) can be considered as a favourable immobilization mechanism in a long-term, leading to precipitation due to the low solubility of U(IV) species. The cast iron insert of the disposal canister and its anaerobic corrosion products are the most important reductants under disposal conditions, but dissolved ferrous iron may also function as reductant. Other iron sources in the buffer or near-field rock, are also considered as possible reductants. The reduction of uranium is a very challenging phenomenon to investigate. The experimental studies need e.g. well-controlled anoxic conditions and measurements of oxidation states. Reduction and other simultaneous phenomena are difficult to distinghuish. The groundwater conditions (pH, Eh and ions) influence on the prevailing complexes of U and Fe and on forming corrosion products of iron and, thus they determine also the redox chemistry. The partial reduction of

  8. Behavior of iodine in the dissolution of spent nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Tsutomu; Komatsu, Kazunori; Takahashi, A. [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-08-01

    The results of laboratory-scale experiments concerning the behavior of iodine in the dissolution of spent nuclear fuels, which were carried out at the Japan Atomic Energy Research Institute, are summarized. Based on previous and new experimental results, the difference in quantity of residual iodine in the fuel solution between laboratory-scale experiments and reprocessing plants is discussed, Iodine in spent fuels is converted to the following four states: (1) oxidation into I{sub 2} by nitric acid, (2) oxidation into I{sub 2} by nitrous acid generated in the dissolution, (3) formation of a colloid of insoluble iodides such as AgI and PdI{sub 2}, and (4) deposition on insoluble residue. Nitrous acid controls the amount of colloid formed. As a result, up to 10% of iodine in spent fuels is retained in the fuel solution, up to 3% is deposited on insoluble residue, and the balance volatilizes to the off-gas, Contrary to earlier belief, when the dissolution is carried out in 3 to 4 M HNO{sub 3} at 100{degrees}C, the main iodine species in a fuel solution is a colloid, not iodate, Immediately after its formation, the colloid is unstable and decomposes partially in the hot nitric acid solution through the following reaction: AgI(s) + 2HNO{sub 3}(aq) = {1/2}I{sub 2}(aq) + AgNO{sub 3}(aq) + NO{sub 2}(g) + H{sub 2}O(1). For high concentrations of gaseous iodine, I{sub 2}(g), and NO{sub 2}, this reaction is reversed towards formation of the colloid (AgI). Since these concentrations are high near the liquid surface of a plant-scale dissolver, there is a possibility that the colloid is formed there through this reversal, Simulations performed in laboratory-scale experiments demonstrated this reversal, This phenomenon can be one reason the quantity of residual iodine in spent fuels is higher in reprocessing plants than in laboratory-scale experiments. 17 refs., 5 figs., 3 tabs.

  9. SSI's standpoints regarding the final disposal of spent nuclear fuel and nuclear waste

    International Nuclear Information System (INIS)

    The report gives an overview of the opinions and statements expressed by the SSI concerning the methodology and the site selection criteria suggested for disposal of spent nuclear fuels in Sweden. The review covers the period from the mid-seventies to 1998

  10. Advantages on dry interim storage for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Romanato, L.S. [Centro Tecnologico da Marinha em Sao Paulo, Av. Professor Lineu Prestes 2468, 05508-900 Sao Paulo (Brazil); Rzyski, B.M. [IPEN/ CNEN-SP, 05508-000 Sao Paulo (Brazil)]. e-mail: romanato@ctmsp.mar.mil.br

    2006-07-01

    When the nuclear fuel lose its ability to efficiently create energy it is removed from the core reactor and moved to a storage unit waiting for a final destination. Generally, the spent nuclear fuel (SNF) remains inside concrete basins with water within the reactors facility for the radioactive activity decay. Water cools the generated heat and shields radioactivity emissions. After some period of time in water basins the SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing installations, or still wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet installations, depending on the method adopted by the nuclear power plant or other plans of the country. In many SNF wet storage sites the capacity can be fulfilled very quickly. If so, additional area or other alternative storage system should be given. There are many options to provide capacity increase in the wet storage area, but dry storages are worldwide preferred since it reduces corrosion concerns. In the wet storage the temperature and water purity should be constantly controlled whereas in the dry storage the SNF stands protected in specially designed canisters. Dry interim storages are practical and approved in many countries especially that have the 'wait and see' philosophy (wait to see new technologies development). This paper shows the advantages of dry interim storages sites in comparison with the wet ones and the nowadays problems as terrorism. (Author)

  11. Final disposal of high levels waste and spent nuclear fuel

    International Nuclear Information System (INIS)

    Foreign and international activities on the final disposal of high-level waste and spent nuclear fuel have been reviewed. A considerable research effort is devoted to development of acceptable disposal options. The different technical concepts presently under study are described in the report. Numerous studies have been made in many countries of the potential risks to future generations from radioactive wastes in underground disposal repositories. In the report the safety assessment studies and existing performance criteria for geological disposal are briefly discussed. The studies that are being made in Canada, the United States, France and Switzerland are the most interesting for Sweden as these countries also are considering disposal into crystalline rocks. The overall time-tables in different countries for realisation of the final disposal are rather similar. Normally actual large-scale disposal operations for high-level wastes are not foreseen until after year 2000. In the United States the Congress recently passed the important Nuclear Waste Policy Act. It gives a rather firm timetable for site-selection and construction of nuclear waste disposal facilities. According to this act the first repository for disposal of commercial high-level waste must be in operation not later than in January 1998. (Author)

  12. Storage facilities of spent nuclear fuel in dry for Mexican nuclear facilities

    International Nuclear Information System (INIS)

    In this article the relevant aspects of the spent fuel storage and the questions that should be taken in consideration for the possible future facilities of this type in the country are approached. A brief description is proposed about the characteristics of the storage systems in dry, the incorporate regulations to the present Nuclear Regulator Standard, the planning process of an installation, besides the approaches considered once resolved the use of these systems; as the modifications to the system, the authorization periods for the storage, the type of materials to store and the consequent environmental impact to their installation. At the present time the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) considers the possible generation of two authorization types for these facilities: Specific, directed to establish a new nuclear installation with the authorization of receiving, to transfer and to possess spent fuel and other materials for their storage; and General, focused to those holders that have an operation license of a reactor that allows them the storage of the nuclear fuel and other materials that they possess. Both authorizations should be valued according to the necessities that are presented. In general, this installation type represents a viable solution for the administration of the spent fuel and other materials that require of a temporary solution previous to its final disposal. Its use in the nuclear industry has been increased in the last years demonstrating to be appropriate and feasible without having a significant impact to the health, public safety and the environment. Mexico has two main nuclear facilities, the nuclear power plant of Laguna Verde of the Comision Federal de Electricidad (CFE) and the facilities of the TRIGA Reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) that will require in a future to use this type of disposition installation of the spent fuel and generated wastes. (Author)

  13. Managing Ageing in Spent Nuclear Fuel Storage Facilities

    International Nuclear Information System (INIS)

    Spent fuel pools (SFP) that are outside containment system without redundancy whose failure could release radioactive material that exceed allowable limit. If SFP have to continue to operate for long term after power plant shutdown it is essential to develop an ageing management program within the general life management program of the nuclear power plant. This work refers to the Atucha I nuclear power plant (NPP) SFPs. The fuel assembly (FA) of Atucha NPPs is 6 meter long and encompasses 36 Zircaloy-4 cladded fuel rods. For these spent fuel assemblies (SFA) there are two storage buildings located adjacent to the reactor building. One of the alternatives considered at the end of Atucha I operation is to transfer all SFAs to dry storage, another one is to continue the operation of the SFPs and to transfer to dry storage just a selected amount of SFAs. For the selection of the dry technology it should be kept in mind the characteristics of the Aturcha SFA, in particular, its length and burnup which differs according to the discharge date because of the use of natural uranium (NU) or slightly enriched uranium (SEU). Therefore, the fundamental point here is to keep in mind that it is the effect of ageing due to time and use that cause net changes in the characteristics of a System, Structure and Component (SSC). We employ formal processes to systematically identify and evaluate the Critical Systems, Structures and Components (CSSCs) in the facilities. A Technology Watch Programme is being established to ensure that degradation mechanisms, which could impact on facilities life, are promptly investigated so that mitigating programmes can be designed. With this methodology we analyse the following components of the pools, concrete wall stability, integrity of concrete structure, pool lining, and integrity of metal structure, pipe failures, degradation in storage racks and SFA degradation. (author)

  14. Radiation induced corrosion of copper for spent nuclear fuel storage

    International Nuclear Information System (INIS)

    The long term safety of repositories for radioactive waste is one of the main concerns for countries utilizing nuclear power. The integrity of engineered and natural barriers in such repositories must be carefully evaluated in order to minimize the release of radionuclides to the biosphere. One of the most developed concepts of long term storage of spent nuclear fuel is the Swedish KBS-3 method. According to this method, the spent fuel will be sealed inside copper canisters surrounded by bentonite clay and placed 500 m down in stable bedrock. Despite the importance of the process of radiation induced corrosion of copper, relatively few studies have been reported. In this work the effect of the total gamma dose on radiation induced corrosion of copper in anoxic pure water has been studied experimentally. Copper samples submerged in water were exposed to a series of total doses using three different dose rates. Unirradiated samples were used as reference samples throughout. The copper surfaces were examined qualitatively using IRAS and XPS and quantitatively using cathodic reduction. The concentration of copper in solution after irradiation was measured using ICP-AES. The influence of aqueous radiation chemistry on the corrosion process was evaluated based on numerical simulations. The experiments show that the dissolution as well as the oxide layer thickness increase upon radiation. Interestingly, the evaluation using numerical simulations indicates that aqueous radiation chemistry is not the only process driving the corrosion of copper in these systems. - Highlights: • Copper cubes were exposed to gamma radiation in anoxic pure water. • The dissolution of copper increases with increasing absorbed total dose. • The oxide layer formed consists mainly of cuprite. • Numerical simulations of the irradiation experiments were performed. • There is a large discrepancy between numerical simulations and experimental results

  15. Survey of experience with dry storage of spent nuclear fuel and update of wet storage experience

    International Nuclear Information System (INIS)

    Spent fuel storage is an important part of spent fuel management. At present about 45,000 t of spent water reactor fuel have been discharged worldwide. Only a small fraction of this fuel (approximately 7%) has been reprocessed. The amount of spent fuel arisings will increase significantly in the next 15 years. Estimates indicate that up to the year 2000 about 200,000 t HM of spent fuel could be accumulated. In view of the large quantities of spent fuel discharged from nuclear power plants and future expected discharges, many countries are involved in the construction of facilities for the storage of spent fuel and in the development of effective methods for spent fuel surveillance and monitoring to ensure that reliable and safe operation of storage facilities is achievable until the time when the final disposal of spent fuel or high level wastes is feasible. The first demonstrations of final disposal are not expected before the years 2000-2020. This is why the long term storage of spent fuel and HLW is a vital problem for all countries with nuclear power programmes. The present survey contains data on dry storage and recent information on wet storage, transportation, rod consolidation, etc. The main aim is to provide spent fuel management policy making organizations, designers, scientists and spent fuel storage facility operators with the latest information on spent fuel storage technology under dry and wet conditions and on innovations in this field. Refs, figs and tabs

  16. Repository for spent nuclear fuel. Plant description layout D - Forsmark

    International Nuclear Information System (INIS)

    This document describes the final repository for spent nuclear fuel, SFK, which is located at Forsmark, in Oesthammar. The bedrock at the site is part of a so-called tectonic lens, in which the rock composition is relatively homogeneous and less deformed than outside the lens. The bedrock consists mainly of granite with high quartz content and good thermal conductivity. The central parts above ground are grouped in an operations area, located at the Soederviken on the south side of the intake duct for cooling water for nuclear power plant. Operating area is divided into an internal, secured portion, where the canisters of fuel are handled and there are links to the underground part, and a outer part, where the buffer, backfill and sealing used in the repository's barriers are produced. The above-ground part of the plant and also include storage of excavated rock, ventilation stations, and supplies of bentonite. The underground portion consists of a central area and a storage area. Caverns of the central area contain features for the underground operation. It communicates with the internal operating range above ground via a spiral ramp and several shafts. The ramp used to transport capsules of spent fuel and other heavy or bulky transport. The shafts are used to transport rock, buffer, backfill and staff, as well as for ventilation. The largest part of the space below ground is the repository where the canisters with the spent fuel are disposed. The capsules are deposited in vertical holes in the tunnels. When the deposit in a tunnel is complete, the tunnel is re-filled. The two main activities underground is rock work and disposal work, which are conducted separately from each other. Rock works covers all steps required to excavate tunnels and drill deposition holes, as well as to make temporary installations in the tunnels. To the landfill works count, besides the deposit of the capsule, the placement of the bentonite buffer in the deposition hole and backfilling

  17. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    International Nuclear Information System (INIS)

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  18. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    Science.gov (United States)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-01

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  19. An overview of burnup credit application in spent nuclear fuel management

    International Nuclear Information System (INIS)

    The current status of burnup credit application has been overviewed for spent nuclear fuel management. It was revealed that the use of burnup credit is practically limited to spent nuclear fuel storage, for which selected actinides-only are taken into account

  20. 78 FR 56775 - Waste Confidence-Continued Storage of Spent Nuclear Fuel

    Science.gov (United States)

    2013-09-13

    ... reactor operating license to dispose of existing commercial high-level radioactive waste and spent nuclear... dispose of the commercial high-level radioactive waste and spent nuclear fuel originating in such reactor... fuel from one high-temperature gas-cooled reactor (HTGR) built and commercially operated: Fort...

  1. Data module development for spent nuclear fuel transportation risk assessment

    International Nuclear Information System (INIS)

    In 1986, in compliance with the Nuclear Waste Policy Act of 1982, the U.S. Department of Energy (DOE) issued environmental assessments (EAs) of the potential repository sites for spent nuclear fuel (SNF) and high-level radioactive wastes. A major concern expressed in the public comments on the EAs was the need for a route-specific risk analysis. One approach to this concern was presented in 1987 by DOE's Office of Civilian Radioactive Waste Management (OCRWM) at a workshop on models for OCRWM transportation risk analysis in Salt Lake City, Utah. The DOE decided that analytical capabilities for a route-specific risk analysis should be maintained at the state level, thus requiring state-specific accident and agricultural output data. The DOE also decided that such an analysis warranted development of a more reliable data base for a number of key input parameters for transportation risk assessment. Such a data base would be compiled in a format acceptable for input into the existing transportation risk code RADTRAN. Seven data modules have been developed on the basis of these considerations. A system module was also developed for integrating the data modules and serves as a preprocessor of the RADTRAN input data file. The relationship of the modules to the RADTRAN input is shown

  2. Final disposal of spent nuclear fuel in the Finnish bedrock

    International Nuclear Information System (INIS)

    Teollisuuden Voima Oy (TVO) studies Finnish bedrock for the final disposal of the spent nuclear fuel from the Olkiluoto nuclear power plant. The study is in accordance with the decision in principle by Finnish government in 1983. The report is the summary of the preliminary site investigations carried out during the years 1987-1992. On the basis of these investigations a few areas will be selected for detailed site investigation. The characterization comprises five areas selected from the shortlist of potential candidate areas resulted in the earlier study during 1983-1985. Areas are located in different parts of Finland and they represent the main formations of the Finnish bedrock. Romuvaara area in Kuhmo and Veitsivaara area in Hyrynsalmi represent the Archean basement. Kivetty area in Konginkangas consists of mainly younger granitic rocks. Syyry in Sievi is located in transition area of Svecofennidic rocks and granitic rocks. Olkiluoto in Eurajoki represents migmatites in southern Finland. For the field investigations area-specific programs were planned and executed. The field investigations have comprised airborne survey by helicopter, geophysical surveys, geological mappings and samplings, deep and shallow core drillings, geophysical and hydrological borehole measurements and groundwater samplings

  3. Multi-Detector Analysis System for Spent Nuclear Fuel Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Reber, Edward Lawrence; Aryaeinejad, Rahmat; Cole, Jerald Donald; Drigert, Mark William; Jewell, James Keith; Egger, Ann Elizabeth; Cordes, Gail Adele

    1999-09-01

    The Spent Nuclear Fuel (SNF) Non-Destructive Analysis (NDA) program at INEEL is developing a system to characterize SNF for fissile mass, radiation source term, and fissile isotopic content. The system is based on the integration of the Fission Assay Tomography System (FATS) and the Gamma-Neutron Analysis Technique (GNAT) developed under programs supported by the DOE Office of Non-proliferation and National Security. Both FATS and GNAT were developed as separate systems to provide information on the location of special nuclear material in weapons configuration (FATS role), and to measure isotopic ratios of fissile material to determine if the material was from a weapon (GNAT role). FATS is capable of not only determining the presence and location of fissile material but also the quantity of fissile material present to within 50%. GNAT determines the ratios of the fissile and fissionable material by coincidence methods that allow the two prompt (immediately) produced fission fragments to be identified. Therefore, from the combination of FATS and GNAT, MDAS is able to measure the fissile material, radiation source term, and fissile isotopics content.

  4. Estimated consequences from severe spent nuclear fuel transportation accidents

    International Nuclear Information System (INIS)

    The RISKIND software package is used to estimate radiological consequences of severe accident scenarios involving the transportation of spent nuclear fuel. Radiological risks are estimated for both a collective population and a maximally exposed individual based on representative truck and rail cask designs described in the U.S. Nuclear Regulatory Commission (NRC) modal study. The estimate of collective population risk considers all possible environmental pathways, including acute and long-term exposures, and is presented in terms of the 50-y committed effective dose equivalent. Radiological risks to a maximally exposed individual from acute exposure are estimated and presented in terms of the first year and 50-y committed effective dose equivalent. Consequences are estimated for accidents occurring in rural and urban population areas. The modeled pathways include inhalation during initial passing of the radioactive cloud, external exposure from a reduction of the cask shielding, long-term external exposure. from ground deposition, and ingestion from contaminated food (rural only). The major pathways and contributing radionuclides are identified, and the effects of possible mitigative actions are discussed. The cask accident responses and the radionuclide release fractions are modeled as described in the NRC modal study. Estimates of severe accident probabilities are presented for both truck and rail modes of transport. The assumptions made in this study tend to be conservative; however, a set of multiplicative factors are identified that can be applied to estimate more realistic conditions

  5. Applying fast calorimetry on a spent nuclear fuel calorimeter

    Energy Technology Data Exchange (ETDEWEB)

    Liljenfeldt, Henrik [Swedish Nuclear Fuel and Waste Management (Sweden); Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Uppsala Univ. (Sweden)

    2015-04-15

    Recently at Los Alamos National Laboratory, sophisticated prediction algorithms have been considered for the use of calorimetry for treaty verification. These algorithms aim to predict the equilibrium temperature based on early data and therefore be able to shorten the measurement time while maintaining good accuracy. The algorithms have been implemented in MATLAB and applied on existing equilibrium measurements from a spent nuclear fuel calorimeter located at the Swedish nuclear fuel interim storage facility. The results show significant improvements in measurement time in the order of 15 to 50 compared to equilibrium measurements, but cannot predict the heat accurately in less time than the currently used temperature increase method can. This Is both due to uncertainties in the calibration of the method as well as identified design features of the calorimeter that limits the usefulness of equilibrium type measurements. The conclusions of these findings are discussed, and suggestions of both improvements of the current calorimeter as well as what to keep in mind in a new design are given.

  6. Spent nuclear fuel transportation: Public issues and answers

    International Nuclear Information System (INIS)

    The court-ordered shipping of 750 spent nuclear fuel assemblies from West Valley, New York back to their utility owners has generated considerable public and media interest. This paper discusses the specific concerns of the general public over the West Valley shipments, the issues raised by opposition groups, the interests of public officials and emergency preparedness teams as well as the media coverage generated. An analysis is performed on the effectiveness of the West Valley and utility public information programs utilized in addressing these issues, concerns and interests. Emphasis is placed on communications which work to facilitate the shipments and generate fuel transport acceptance. Information programs are discussed which increase preparedness for nuclear shipments by emergency response teams and build public confidence in their safety. The paper also examines communications which could have further enhanced the shipping campaigns to date. Finally, plans are discussed for media preparation with interview training and press conferences. Emphasis is placed on materials provided for the media which has served to generate more favorable print and air time

  7. Managing Spent Nuclear Fuel at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Thomas Hill; Denzel L. Fillmore

    2005-10-01

    The Idaho National Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy derives from the history of the INL as the National Reactor Testing Station, and from its mission to recover HEU from SNF and to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facilities, some 50 years old. SNF at INL has many forms—from intact assemblies down to metallurgical mounts, and some fuel has been wet stored for over 40 years. SNF is stored bare or in metal cans under water, or dry in vaults, caissons or casks. Inspection shows varying corrosion and degradation of the SNF and its storage cans. SNF has been stored in 10 different facilities: 5 pools, one cask storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The pools range in age from 40 years old to the most modern in the US Department of Energy (DOE) complex. The near-term objective is to move SNF from older pools to interim dry storage, allowing shutdown and decommissioning of the older facilities. This move involves drying methods that are dependent on fuel type. The long-term objective is to have INL SNF in safe dry storage and ready to be shipped to the National Repository. The unique features of the INL SNF requires special treatments and packaging to meet the proposed repository acceptance criteria and SNF will be repackaged in standardized canisters for shipment and disposal in the National Repository. Disposal will use the standardized canisters that can be co-disposed with High Level Waste glass logs to limit the total fissile material in a repository waste package. The DOE standardized canister also simplifies the repository handling of the multitude of DOE SNF sizes and shapes.

  8. Issues related to EM management of DOE spent nuclear fuel

    International Nuclear Information System (INIS)

    This document is a summary of the important issues involved in managing spent nuclear fuel (SNF) owned by the Department of Energy (DOE). Issues related to civilian SNF activities are not discussed. DOE-owned SNF is stored primarily at the Hanford Site, Idaho National Engineering Laboratory (INEL), Savannah River Site (SRS), Oak Ridge National Laboratory (ORNL), and West Valley Demonstration Project. Smaller quantities of SNF are stored at Brookhaven National Laboratory, Sandia National Laboratories, and Los Alamos National Laboratory (LANL). There is a wide variety of fuel types, including both low and high enrichment fuels from weapons production, DOE reactors, research and development programs, naval programs, and universities. Most fuel is stored in pools associated with reactor or reprocessing facilities. Smaller quantities are in dry storage. Physical conditions of the fuel range from excellent to poor or severely damaged. An issue is defined as an important question that must be answered or decision that must be made on a topic or subject relevant to achieving the complimentary objectives of (a) storing SNF in compliance with applicable regulations and orders until it can be disposed, and (b) safely disposing of DOE's SNF. The purpose of this document is to define the issues; no recommendations are made on resolutions. As DOE's national SNF management program is implemented, a system of issues identification, documentation, tracking, and resolution will be implemented. This document is an initial effort at issues identification. The first section of this document is an overview of issues that are common to several or all DOE facilities that manage SNF. The common issues are organized according to specific aspects of spent fuel management. This is followed by discussions of management issues that apply specifically to individual DOE facilities. The last section provides literature references

  9. Issues related to EM management of DOE spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, D.G. [EG& G Idaho, Inc., Idaho Falls, ID (United States); Abashian, M.S.; Chakraborti, S.; Roberson, K.; Meloin, J.M. [IT Corp. (United States)

    1993-07-01

    This document is a summary of the important issues involved in managing spent nuclear fuel (SNF) owned by the Department of Energy (DOE). Issues related to civilian SNF activities are not discussed. DOE-owned SNF is stored primarily at the Hanford Site, Idaho National Engineering Laboratory (INEL), Savannah River Site (SRS), Oak Ridge National Laboratory (ORNL), and West Valley Demonstration Project. Smaller quantities of SNF are stored at Brookhaven National Laboratory, Sandia National Laboratories, and Los Alamos National Laboratory (LANL). There is a wide variety of fuel types, including both low and high enrichment fuels from weapons production, DOE reactors, research and development programs, naval programs, and universities. Most fuel is stored in pools associated with reactor or reprocessing facilities. Smaller quantities are in dry storage. Physical conditions of the fuel range from excellent to poor or severely damaged. An issue is defined as an important question that must be answered or decision that must be made on a topic or subject relevant to achieving the complimentary objectives of (a) storing SNF in compliance with applicable regulations and orders until it can be disposed, and (b) safely disposing of DOE`s SNF. The purpose of this document is to define the issues; no recommendations are made on resolutions. As DOE`s national SNF management program is implemented, a system of issues identification, documentation, tracking, and resolution will be implemented. This document is an initial effort at issues identification. The first section of this document is an overview of issues that are common to several or all DOE facilities that manage SNF. The common issues are organized according to specific aspects of spent fuel management. This is followed by discussions of management issues that apply specifically to individual DOE facilities. The last section provides literature references.

  10. Thermal analysis of spent nuclear fuel shipping cas

    International Nuclear Information System (INIS)

    In this study, a computational fluid dynamics (CFD) thermal analysis was performed for the TN-24P cask. For the analysis, ANSYS Fluent as a CFD tool was selected since it has the proper finite volume methods to realistically simulate the thermal behavior of shipping casks. For the analysis, spent fuels discharged from pressurized water reactors (PWRs) were modeled. In the model, there are 24 PWR spent fuel assemblies loaded in the TN-24P cask. The fuel design is assumed to be similar to standard Westinghouse 15x15 rod design. Total heat (decay) generated in the cask was estimated to be 20.6 kW. To input the axial power profile required to calculate the heat flux, a User Defined Function was generated. Fuel storage space (canister) is filled with Helium gas to cool spent nuclear fuel. In the cask, heat transfer occurs through the heat conduction by helium and basket, natural circulation driven by gravity, and thermal radiation in the complex geometry. In the canister region, laminar flow model with Boussinesq approximation is used to simulate the natural circulation. The helium domain was assumed symmetric in the model. For thermal radiation, the Discrete Ordinates (DO) model was chosen in the presented study due to its accuracy and capability of parallel processing. In typical vertical TN-24P dry storage cask system consist of two nested cask. Between inner and outer cask is in the air. Air inlet section is at the bottom side of cask and outlet ventilation is at top of cask. At this region, turbulence regime occurs and turbulence is modeled by using k-epsilon model. The analysis include small scaled and full scaled model. In small scale model, geometry is defined rectangular to make mesh generation easy and to validate the analysis tools using the experimental data. In the full-scale simulation, the results of analysis and experimental data for peak clad temperature (PCT) were compared. Key Words: TN-24P dry storage cask, CFD, thermal analysis, PCT, air blockage

  11. Decommissioning and dismantling of nuclear reactors and nuclear spent fuel interim storage in Germany

    International Nuclear Information System (INIS)

    The authors visited Germany in April 2013 to investigate state of reactor decommissioning and dismantling and interim storage of spent fuels reflecting nuclear power phaseout policy after the Fukushima accident. They visited interim storage facilities of radioactive wastes (ZLN, Zwischenlanger Nord) and central active workshop (ZAW, Zentrale Aktive Werkstatt) at Greifswald, and interim storage facilities of spent fuels at Philippsburg. CASTOR (Cask for Storage and Transport of Radioactive Material) was used for interim storage of spent fuels and high-level wastes for 40 years. Amount of wastes produced by decommissioning and dismantling was estimated 1800 ktons consisting of 1200 ktons non-radioactive and 600 ktons radioactive wastes, 500 ktons of which could be decontaminated less than clearance level and 100 ktons of which were obliged to be stored as radioactive wastes. New geological repository site for high level radioactive wastes should be found and developed. (T. Tanaka)

  12. Spent nuclear fuel removal program at the West Valley Demonstration Project: Topical report

    Energy Technology Data Exchange (ETDEWEB)

    Connors, B. J.; Golden, M. P.; Valenti, P. J.; Winkel, J. J.

    1987-03-01

    The spent nuclear fuel removal program at the West Valley Demonstration Project (WVDP) consisted of removing the spent nuclear fuel (SNF) assemblies from the storage pool in the plant, loading them in shielded casks, and preparing the casks for transportation. So far, four fuel removal campaigns have been completed with the return of 625 spent nuclear fuel assemblies to their four utility owners. A fifth campaign, which is not yet completed, will transfer the remaining 125 fuel assemblies to a government site in Idaho. A spent fuel rod consolidation demonstration has been completed, and the storage canisters and their racks are being removed from the fuel receiving and storage pool to make way for installation of the size reduction equipment. A brief history of the West Valley reprocessing plant and the events leading to the storage and ownership of the spent nuclear fuel assemblies and their subsequent removal from West Valley are also recorded as background information. 3 refs., 16 figs., 9 tabs.

  13. Spent nuclear fuel removal program at the West Valley Demonstration Project: Topical report

    International Nuclear Information System (INIS)

    The spent nuclear fuel removal program at the West Valley Demonstration Project (WVDP) consisted of removing the spent nuclear fuel (SNF) assemblies from the storage pool in the plant, loading them in shielded casks, and preparing the casks for transportation. So far, four fuel removal campaigns have been completed with the return of 625 spent nuclear fuel assemblies to their four utility owners. A fifth campaign, which is not yet completed, will transfer the remaining 125 fuel assemblies to a government site in Idaho. A spent fuel rod consolidation demonstration has been completed, and the storage canisters and their racks are being removed from the fuel receiving and storage pool to make way for installation of the size reduction equipment. A brief history of the West Valley reprocessing plant and the events leading to the storage and ownership of the spent nuclear fuel assemblies and their subsequent removal from West Valley are also recorded as background information. 3 refs., 16 figs., 9 tabs

  14. Nuclear spent fuel management scenarios. Status and assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Dufek, J.; Arzhanov, V.; Gudowski, W. [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Nuclear and Reactor Physics

    2006-06-15

    The strategy for management of spent nuclear fuel from the Swedish nuclear power programme is interim storage for cooling and decay for about 30 years followed by direct disposal of the fuel in a geologic repository. In various contexts it is of interest to compare this strategy with other strategies that might be available in the future as a result of ongoing research and development. In particular partitioning and transmutation is one such strategy that is subject to considerable R and D-efforts within the European Union and in other countries with large nuclear programmes. To facilitate such comparisons for the Swedish situation, with a planned phase out of the nuclear power programme, SKB has asked the team at Royal Inst. of Technology to describe and explore some scenarios that might be applied to the Swedish programme. The results of this study are presented in this report. The following scenarios were studied by the help of a specially developed computer programme: Phase out by 2025 with direct disposal. Burning plutonium and minor actinides as MOX in BWR. Burning plutonium and minor actinides as MOX in PWR. Burning plutonium and minor actinides in ADS. Combined LWR-MOX plus ADS. For the different scenarios nuclide inventories, waste amounts, costs, additional electricity production etc have been assessed. As a general conclusion it was found that BWR is more efficient for burning plutonium in MOX fuel than PWR. The difference is approximately 10%. Furthermore the BWR produces about 10% less americium inventory. An ADS reactor park can theoretically in an ideal case burn (transmute) 99% of the transuranium isotopes. The duration of such a scenario heavily depends on the interim time needed for cooling the spent fuel before reprocessing. Assuming 10 years for cooling of nuclear fuel from ADS, the duration will be at least 200 years under optimistic technical assumptions. The development and use of advanced pyro-processing with an interim cooling time of only

  15. A robotized surface workstation for manipulation, filling and closing of packaging containers for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bartos, Pavel [FITE a.s., Ostrava-Marianske Hory (Czech Republic); Haladova, Petra [Robotsystem, LLC/Moravian Research, LLC, Ostrava-Moravska (Czech Republic); Otcenasek, Petr

    2016-01-15

    Options for the handling of spent nuclear fuel are described and a packaging cask for an underground repository is presented as also a robotic surface workplace for the repository. The potential for the closing the nuclear fuel cycle is discussed. Currently, a team of Czech experts is developing a project of fully robotic technology for manipulation and storage of packaging casks for spent nuclear fuel in host rock of underground repository.

  16. A robotized surface workstation for manipulation, filling and closing of packaging containers for spent nuclear fuel

    International Nuclear Information System (INIS)

    Options for the handling of spent nuclear fuel are described and a packaging cask for an underground repository is presented as also a robotic surface workplace for the repository. The potential for the closing the nuclear fuel cycle is discussed. Currently, a team of Czech experts is developing a project of fully robotic technology for manipulation and storage of packaging casks for spent nuclear fuel in host rock of underground repository.

  17. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    Energy Technology Data Exchange (ETDEWEB)

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  18. 76 FR 81542 - In the Matter of ZIONSOLUTIONS, LLC; Zion Nuclear Power Station; Independent Spent Fuel Storage...

    Science.gov (United States)

    2011-12-28

    ... COMMISSION In the Matter of ZIONSOLUTIONS, LLC; Zion Nuclear Power Station; Independent Spent Fuel Storage..., Licensing and Inspection Directorate, Division of Spent Fuel Storage and Transportation, Office of Nuclear... providing notice, in the matter of Zion Nuclear ] Power Station Independent Spent Fuel Storage...

  19. Irradiation of Microbes from Spent Nuclear Fuel Storage Pool Environments

    International Nuclear Information System (INIS)

    Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additional microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 103 and 6 x 104 rad. All microbial specimens withstood the lower exposure with little or no reduction in cell population. Upon exposing the microbes to the larger dose of 6 x 104 rad, we observed two distinct groupings: microbes that demonstrate resistance to radiation, and microbes that display intolerance through a dramatic reduction from their initial population. Microbes in the radiation tolerant grouping were exposed to 1.1 x 105 rad to examine the extent of their resistance. We observe a correlation between radiation resistance and gram stain. The gram-positive species we examined seem to demonstrate a greater radiation resistance

  20. Microbial sampling of aluminum-clad spent nuclear fuel

    International Nuclear Information System (INIS)

    A microbial sampling program was initiated at the Idaho National Engineering and Environmental Laboratory (INEEL) to ascertain the effect of microbial activity on the corrosion of aluminum clad spent nuclear fuel (SNF) stored in wet and dry conditions. In the newest fuel storage pool at the INEEL (CPP-666) pitting corrosion has been observed on aluminum corrosion coupons that can not be explained by the excellent water chemistry. Pitting corrosion of the aluminum-clad SNF and corrosion coupons has been observed in the older fuel storage pool (CPP-603). Therefore a microbial assessment of the bulk water, and basin surfaces of both fuel pools was conducted. The results of this microbial enumeration show that a viable and active microbial population does exist in planktonic form. Sampling of aluminum corrosion coupons placed next to stored fuel elements show that microbial attachment has occurred and a biofilm has formed. The sampling program was then extended to the surfaces of wet and dry stored fuel elements. Viable cells or spores were found on the surfaces of the ATR fuel elements that were stored under wet and dry conditions. This paper discusses the methodology of sampling the surfaces of SNF stored under wet conditions for the presence of microorganisms and the types of organisms found

  1. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    Energy Technology Data Exchange (ETDEWEB)

    N. E. Pettit

    2001-07-13

    The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident.

  2. Dose reduction improvements in storage basins of spent nuclear fuel

    International Nuclear Information System (INIS)

    Spent nuclear fuel in storage basins at the Hanford Site has corroded and contaminated basin water, which has leaked into the soil; the fuel also had deposited a layer of radioactive sludge on basin floors. The SNF is to be removed from the basins to protect the nearby Columbia River. Because the radiation level is high, measures have been taken to reduce the background dose rate to as low as reasonably achievable (ALARA) to prevent radiation doses from becoming the limiting factor for removal of the SW in the basins to long-term dry storage. All activities of the SNF Project require application of ALARA principles for the workers. On the basis of these principles dose reduction improvements have been made by first identifying radiological sources. Principal radiological sources in the basin are basin walls, basin water, recirculation piping and equipment. Dose reduction activities focus on cleaning and coating basin walls to permit raising the water level, hydrolasing piping, and placing lead plates. In addition, the transfer bay floor will be refinished to make decontamination easier and reduce worker exposures in the radiation field. The background dose rates in the basin will be estimated before each task commences and after it is completed; these dose reduction data will provide the basis for cost benefit analysis

  3. Safety Tests of Concrete Storage Cask for Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    In preparation for the timely installation of interim storage facility for spent nuclear fuel (SF), KORAD is developing domestic models of SF storage systems and the concrete storage cask is one of them. A concrete cask consists of a metallic canister which confines SF with welded closure and a concrete overpack which provides radiation shielding and physical protection to the canister. The safety requirements for a SF storage cask is well established in US and summarized in regulatory guides such as NUREG-1536. KAERI has been performing tests of the concrete cask to demonstrate its safety and compliance to the regulatory requirements with high priority stipulated in NUREG-1536. The test program includes the structural performance tests under tip-over and earthquake and decay heat removal test under normal, off-normal and accident conditions. In this paper, brief introduction to the structural tests and their results are provided. Safety tests to demonstrate the safety of KORAD21C concrete storage cask were successfully performed. The structural integrity during tip-over and earthquake were demonstrated with scale model tests and the results are analyzed in comparison with safety analysis results

  4. Spent Nuclear Fuel (SNF) Project Design Verification and Validation Process

    International Nuclear Information System (INIS)

    This document provides a description of design verification and validation activities implemented by the Spent Nuclear Fuel (SNF) Project. During the execution of early design verification, a management assessment (Bergman, 1999) and external assessments on configuration management (Augustenburg, 1999) and testing (Loscoe, 2000) were conducted and identified potential uncertainties in the verification process. This led the SNF Chief Engineer to implement corrective actions to improve process and design products. This included Design Verification Reports (DVRs) for each subproject, validation assessments for testing, and verification of the safety function of systems and components identified in the Safety Equipment List to ensure that the design outputs were compliant with the SNF Technical Requirements. Although some activities are still in progress, the results of the DVR and associated validation assessments indicate that Project requirements for design verification are being effectively implemented. These results have been documented in subproject-specific technical documents (Table 2). Identified punch-list items are being dispositioned by the Project. As these remaining items are closed, the technical reports (Table 2) will be revised and reissued to document the results of this work

  5. Spent Nuclear Fuel (SNF) Project Design Verification and Validation Process

    Energy Technology Data Exchange (ETDEWEB)

    OLGUIN, L.J.

    2000-09-25

    This document provides a description of design verification and validation activities implemented by the Spent Nuclear Fuel (SNF) Project. During the execution of early design verification, a management assessment (Bergman, 1999) and external assessments on configuration management (Augustenburg, 1999) and testing (Loscoe, 2000) were conducted and identified potential uncertainties in the verification process. This led the SNF Chief Engineer to implement corrective actions to improve process and design products. This included Design Verification Reports (DVRs) for each subproject, validation assessments for testing, and verification of the safety function of systems and components identified in the Safety Equipment List to ensure that the design outputs were compliant with the SNF Technical Requirements. Although some activities are still in progress, the results of the DVR and associated validation assessments indicate that Project requirements for design verification are being effectively implemented. These results have been documented in subproject-specific technical documents (Table 2). Identified punch-list items are being dispositioned by the Project. As these remaining items are closed, the technical reports (Table 2) will be revised and reissued to document the results of this work.

  6. An approach to optimizing spent nuclear fuel transportation logistics

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE) has been assigned responsibility for managing the nation's spent nuclear fuel (SNF). Because of the large quantity of SNF to be moved from the reactors in which it is generated, the DOE is planning a new generation of transport casks designed specifically for the characteristics of the SNF to be transported. Choosing the age and burnup values for which to design the transportation cask would be relatively simple if all the SNF to be transported in a new generation of casks were expected to have the same age and burnup characteristics. It is desirable to maximize cask capacity, which leads to fewer cask trips and thus to less cost and less public risk. One could design multiple casks, each tailored to a specific portion of the SNF, and reduce the total number of trips needed to transport a fixed quantity of SNF. However, the additional administrative, design, and certification costs could outweigh any savings accruing from reduced trips. a significant part of ongoing transportation systems analysis in the U.S. is devoted to the design issues described above. This paper outlines some of the approaches currently under consideration

  7. Irradiation of Microbes from Spent Nuclear Fuel Storage Pool Environments

    Energy Technology Data Exchange (ETDEWEB)

    Breckenridge, C.R.; Watkins, C.S.; Bruhn, D.F.; Roberto, F.F.; Tsang, M.N.; Pinhero, P.J. [INEEL (US); Brey, R.F. [ISU (US); Wright, R.N.; Windes, W.F.

    1999-09-03

    Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additional microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 10{sup 3} and 6 x 10{sup 4} rad. All microbial specimens withstood the lower exposure with little or no reduction in cell population. Upon exposing the microbes to the larger dose of 6 x 10{sup 4} rad, we observed two distinct groupings: microbes that demonstrate resistance to radiation, and microbes that display intolerance through a dramatic reduction from their initial population. Microbes in the radiation tolerant grouping were exposed to 1.1 x 10{sup 5} rad to examine the extent of their resistance. We observe a correlation between radiation resistance and gram stain. The gram-positive species we examined seem to demonstrate a greater radiation resistance.

  8. High Burn-Up Spent Nuclear Fuel Vibration Integrity Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Rob L [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    The Oak Ridge National Laboratory (ORNL) has developed the cyclic integrated reversible-bending fatigue tester (CIRFT) approach to successfully demonstrate the controllable fatigue fracture on high burnup (HBU) spent nuclear fuel (SNF) in a normal vibration mode. CIRFT enables examination of the underlying mechanisms of SNF system dynamic performance. Due to the inhomogeneous composite structure of the SNF system, the detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained from a CIRFT system measurement. Therefore, finite element analyses (FEAs) are used to translate the global moment-curvature measurement into local stress-strain profiles for further investigation. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance. Fuel structure contributes to SNF system stiffness. There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interactions. SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.

  9. Fission Product Release from Spent Nuclear Fuel During Melting

    International Nuclear Information System (INIS)

    The Melt-Dilute process consolidates aluminum-clad spent nuclear fuel by melting the fuel assemblies and diluting the 235U content with depleted uranium to lower the enrichment. During the process, radioactive fission products whose boiling points are near the proposed 850 degrees C melting temperature can be released. This paper presents a review of fission product release data from uranium-aluminum alloy fuel developed from Severe Accident studies. In addition, scoping calculations using the ORIGEN-S computer code were made to estimate the radioactive inventories in typical research reactor fuel as a function of burnup, initial enrichment, and reactor operating history and shutdown time.Ten elements were identified from the inventory with boiling points below or near the 850 degrees C reference melting temperature. The isotopes 137Cs and 85Kr were considered most important. This review serves as basic data to the design and development of a furnace off-gas system for containment of the volatile species

  10. Refinishing contamination floors in Spent Nuclear Fuels storage basins

    International Nuclear Information System (INIS)

    The floors of the K Basins at the Hanford Site are refinished to make decontamination easier if spills occur as the spent nuclear fuel (SNF) is being unloaded from the basins for shipment to dry storage. Without removing the contaminated existing coating, the basin floors are to be coated with an epoxy coating material selected on the basis of the results of field tests of several paint products. The floor refinishing activities must be reviewed by a management review board to ensure that work can be performed in a controlled manner. Major documents prepared for management board review include a report on maintaining radiation exposure as low as reasonably achievable, a waste management plan, and reports on hazard classification and unreviewed safety questions. To protect personnel working in the radiation zone, Operational Health Physics prescribed the required minimum protective methods and devices in the radiological work permit. Also, industrial hygiene safety must be analyzed to establish respirator requirements for persons working in the basins. The procedure and requirements for the refinishing work are detailed in a work package approved by all safety engineers. After the refinishing work is completed, waste materials generated from the refinishing work must be disposed of according to the waste management plan

  11. Hanford K basins spent nuclear fuel project update

    International Nuclear Information System (INIS)

    Twenty one hundred metric tons of spent nuclear fuel (SNF) are currently stored in the Hanford Site K Basins near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported to the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building

  12. Nuclear fuel burn-up credit for criticality safety justification of spent nuclear fuel storage systems

    International Nuclear Information System (INIS)

    Burn-up credit analysis of RBMK-1000 an WWER-1000 spent nuclear fuel accounting only for actinides is carried out and a method is proposed for actinide burn-up credit. Two burn-up credit approaches are analyzed, which consider a system without and with the distribution of isotopes along the height of the fuel assembly. Calculations are performed using SCALE and MCNP computer codes

  13. INEL integrated spent nuclear fuel consolidation task team report

    Energy Technology Data Exchange (ETDEWEB)

    Henry, R.N.; Clark, J.H.; Chipman, N.A. [and others

    1994-09-12

    This document describes a draft plan and schedule to consolidate spent nuclear fuel (SNF) and special nuclear material (SNW) from aging storage facilities throughout the Idaho National Engineering Laboratory (INEL) to the Idaho Chemical Processing Plant (ICPP) in a safe, cost-effective, and expedient manner. A fully integrated and resource-loaded schedule was developed to achieve consolidation as soon as possible. All of the INEL SNF and SNM management task, projects, and related activities from fiscal year 1994 to the end of the consolidation period are logic-tied and integrated with each other. The schedule and plan are presented to initiate discussion of their implementation, which is expected to generate alternate concepts that can be evaluated using the methodology described in this report. Three perturbations to consolidating SNF as soon as possible are also explored. If the schedule is executed as proposed, the new and on-going consolidation activities will require about 6 years to complete and about $25.3M of additional funding. Reduced annual operating costs are expected to recover the additional investment in about 6.4 years. The total consolidation program as proposed will cost about $66.8M and require about 6 years to recover via reduced operating costs from retired SNF/SNM storage facilities. Detailed schedules and cost estimates for the Test Reactor Area Materials Test Reactor canal transfers are included as an example of the level of detail that is typical of the entire schedule (see Appendix D). The remaining work packages for each of the INEL SNF consolidation transfers are summarized in this document. Detailed cost and resource information is available upon request for any of the SNF consolidation transfers.

  14. Spent nuclear fuel storage. (Latest citations from the NTIS bibliographic database). Published Search

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    The bibliography contains citations concerning spent nuclear fuel storage technologies, facilities, sites, and assessment. References review wet and dry storage, spent fuel casks and pools, underground storage, monitored and retrievable storage systems, and aluminum-clad spent fuels. Environmental impact, siting criteria, regulations, and risk assessment are also discussed. Computer codes and models for storage safety are covered. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  15. On-site interim storage of spent nuclear fuel: Emerging public issues

    International Nuclear Information System (INIS)

    Failure to consummate plans for a permanent repository or above- ground interim Monitored Retrievable Storage (MRS) facility for spent nuclear fuel has spurred innovative efforts to ensure at-reactor storage in an environmentally safe and secure manner. This article examines the institutional and socioeconomic impacts of Dry Cask Storage Technology (DCST)-an approach to spent fuel management that is emerging as the preferred method of on-site interim spent fuel storage by utilities that exhaust existing storage capacity

  16. The Shipment of Russian-Origin Highly Enriched Uranium Research Reactor Spent Nuclear Fuel from Belarus

    Energy Technology Data Exchange (ETDEWEB)

    Sikorin, S.N.; Polazau, S.A.; Hryharovich, T.K. [Joint Institute for Power and Nuclear Research ' Sosny' , Academik Krasin Street, Minsk (Belarus); Bolshinsky, I. [Idaho National Laboratory, N. Fremont Avenue Idaho Falls, Idaho (United States); Thomas, J.E. [Savannah River National Laboratory, Aiken, South Carolina (United States)

    2011-07-01

    In October 2010, the Global Threat Reduction Initiative and the Joint Institute for Power and Nuclear Research - 'Sosny' of the National Academy of Sciences of the Republic of Belarus completed a shipment that returned 43 kilograms of Russian-origin highly enriched uranium (HEU) spent nuclear fuel to the Russian Federation. The spent fuel was legacy material, discharged from the two decommissioned reactors, the Pamir-630D mobile reactor and the IRT-M research reactor. This shipment marked the complete removal of all HEU spent nuclear fuel from Belarus. This paper discusses the planning, preparations, and coordination required to complete this important international shipment successfully. (author)

  17. The Shipment of Russian-Origin Highly Enriched Uranium Research Reactor Spent Nuclear Fuel from Belarus

    International Nuclear Information System (INIS)

    In October 2010, the Global Threat Reduction Initiative and the Joint Institute for Power and Nuclear Research - 'Sosny' of the National Academy of Sciences of the Republic of Belarus completed a shipment that returned 43 kilograms of Russian-origin highly enriched uranium (HEU) spent nuclear fuel to the Russian Federation. The spent fuel was legacy material, discharged from the two decommissioned reactors, the Pamir-630D mobile reactor and the IRT-M research reactor. This shipment marked the complete removal of all HEU spent nuclear fuel from Belarus. This paper discusses the planning, preparations, and coordination required to complete this important international shipment successfully. (author)

  18. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix D, Part B: Naval spent nuclear fuel management

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    This volume contains the following attachments: transportation of Naval spent nuclear fuel; description of Naval spent nuclear receipt and handling at the Expended Core Facility at the Idaho National Engineering Laboratory; comparison of storage in new water pools versus dry container storage; description of storage of Naval spent nuclear fuel at servicing locations; description of receipt, handling, and examination of Naval spent nuclear fuel at alternate DOE facilities; analysis of normal operations and accident conditions; and comparison of the Naval spent nuclear fuel storage environmental assessment and this environmental impact statement.

  19. U.S. Nuclear Regulatory Commission's package performance study for spent nuclear fuel transportation

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission (NRC) is currently planning to conduct confirmatory analyses and testing of the performance of Type B spent nuclear fuel (SNF) packages during severe accidents. Current plans call for full-scale rail impact and thermal tests, with pre-test predictions, of a truck cask design (∼50 ton) and a rail cask design (∼140 ton), in late 2004-2005. A public participatory process is being used to design and conduct this project, with a principle objective of increasing public confidence. (author)

  20. Direct Investigations of the Immobilization of Radionuclides in the Alteration Products of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Peter C. Burns; Robert J. Finch; David J. Wronkiewicz

    2004-12-27

    Safe disposal of the nation's nuclear waste in a geological repository involves unique scientific and engineering challenges owing to the very long-lived radioactivity of the waste. The repository must retain a variety of radionuclides that have vastly different chemical characters for several thousand years. Most of the radioactivity that will be housed in the proposed repository at Yucca Mountain will be associated with spent nuclear fuel, much of which is derived from commercial reactors. DOE is custodian of approximately 8000 tons of spent nuclear fuel that is also intended for eventual disposal in a geological repository. Unlike the spent fuel from commercial reactors, the DOE fuel is diverse in composition with more than 250 varieties. Safe disposal of spent fuel requires a detailed knowledge of its long-term behavior under repository conditions, as well as the fate of radionuclides released from the spent fuel as waste containers are breached.

  1. Direct Investigations of the Immobilization of Radionuclides in the Alteration Products of Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Safe disposal of the nation's nuclear waste in a geological repository involves unique scientific and engineering challenges owing to the very long-lived radioactivity of the waste. The repository must retain a variety of radionuclides that have vastly different chemical characters for several thousand years. Most of the radioactivity that will be housed in the proposed repository at Yucca Mountain will be associated with spent nuclear fuel, much of which is derived from commercial reactors. DOE is custodian of approximately 8000 tons of spent nuclear fuel that is also intended for eventual disposal in a geological repository. Unlike the spent fuel from commercial reactors, the DOE fuel is diverse in composition with more than 250 varieties. Safe disposal of spent fuel requires a detailed knowledge of its long-term behavior under repository conditions, as well as the fate of radionuclides released from the spent fuel as waste containers are breached

  2. RADIOLYTIC AND THERMAL PROCESSES RELEVANT TO DRY STORAGE OF SPENT NUCLEAR FUELS

    Science.gov (United States)

    Thousands of tons of metallic uranium spent-nuclear-fuel (SNF) remain in water storage across the Department of Energy complex. For example, the Hanford Site K-Basins hold 2300 metric tons of spent fuel, much of it severely corroded. Similar situations exist elsewhere in the DOE ...

  3. Environmental Impact Statement. March 2011. Interim storage, encapsulation and final disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    An Environmental Impact Statement (EIS) shall be prepared and submitted along with applications for permissibility and a licence under the Environmental Code and a licence under the Nuclear Activities Act for new nuclear facilities. This Environmental Impact Statement has been prepared by Svensk Kaernbraenslehantering AB (the Swedish Nuclear Fuel and Waste Management Co, SKB) to be included in the licence applications for continued operation of Clab (central interim storage facility for spent nuclear fuel) in Simpevarp in Oskarshamn Municipality and construction and operation of facilities for encapsulation (integrated with Clab) and final disposal of spent nuclear fuel in Forsmark in Oesthammar Municipality

  4. Environmental Impact Statement. March 2011. Interim storage, encapsulation and final disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    2011-07-01

    An Environmental Impact Statement (EIS) shall be prepared and submitted along with applications for permissibility and a licence under the Environmental Code and a licence under the Nuclear Activities Act for new nuclear facilities. This Environmental Impact Statement has been prepared by Svensk Kaernbraenslehantering AB (the Swedish Nuclear Fuel and Waste Management Co, SKB) to be included in the licence applications for continued operation of Clab (central interim storage facility for spent nuclear fuel) in Simpevarp in Oskarshamn Municipality and construction and operation of facilities for encapsulation (integrated with Clab) and final disposal of spent nuclear fuel in Forsmark in Oesthammar Municipality

  5. Melt-dilute treatment of spent nuclear fuel assemblies from research and test reactors

    International Nuclear Information System (INIS)

    The Savannah River Site is the U.S. Department of Energy's preferred site for return and treatment of all aluminum-base, spent, research and test reactor fuel assemblies. There are over 20,000 spent fuel assemblies now stored in different countries around the world, and by 2035 many will be returned to SRS for treatment and interim storage, in preparation for disposal in a geologic repository. The early fuel assemblies for research and test reactors were made using aluminum clad plates that were fabricated from highly enriched (93%) uranium-aluminum alloy. Later, powder metallurgical fabrication methods were developed to produce plate fuels with higher uranium contents using either uranium aluminide, uranium oxide or uranium silicide powders mixed with aluminum. Silicide fuel elements generally are fabricated with low enriched uranium containing less than 20% 2'35U. Following irradiation, the spent fuel assemblies are discharged from the reactor, and most assemblies have been stored in underwater pools, some since the early 1950's. A number of disposition options including direct/co-disposal and melt-dilute treatment were evaluated recently. The melt-dilute technique was identified as the preferred method for treatment of aluminum-base spent fuel. The technique consists of melting the spent fuel assembly and adding depleted uranium to the melt for isotopic dilution to 2'35U. Aluminum is added, if necessary, to produce a predetermined alloy composition. Additionally, neutron poisons may be added to the melt where they form solid solution phases or compounds with uranium and/or aluminum. Lowering the enrichment reduces both criticality and proliferation concerns for storage. Consolidation by melting also reduces the number of storage canisters. Laboratory and small-scale process demonstration using irradiated fuel is underway. Tests of the off gas absorption system have been initiated using both surrogate and irradiated RERTR mini fuel plates. An experimental L

  6. MANAGING SPENT NUCLEAR FUEL WASTES AT THE IDAHO NATIONAL LABORATORY

    Energy Technology Data Exchange (ETDEWEB)

    Hill, Thomas J

    2005-09-01

    The Idaho National Engineering Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy is in part due to the history of the INL as the National Reactor Testing Station, in part to its mission to recover highly enriched uranium from SNF and in part to it’s mission to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facility, some dating back 50 years in the site history. The success of the INL SNF program is measured by its ability to: 1) achieve safe existing storage, 2) continue to receive SNF from other locations, both foreign and domestic, 3) repackage SNF from wet storage to interim dry storage, and 4) prepare the SNF for dispositioning in a federal repository. Because of the diversity in the SNF and the facilities at the INL, the INL is addressing almost very condition that may exist in the SNF world. Many of solutions developed by the INL are applicable to other SNF storage sites as they develop their management strategy. The SNF being managed by the INL are in a variety of conditions, from intact assemblies to individual rods or plates to powders, rubble, and metallurgical mounts. Some of the fuel has been in wet storage for over forty years. The fuel is stored bare, or in metal cans and either wet under water or dry in vaults, caissons or casks. Inspections have shown varying degrees of corrosion and degradation of the fuel and the storage cans. Some of the fuel has been recanned under water, and the conditions of the fuel inside the second or third can are unknown. The fuel has been stored in one of 10 different facilities: five wet pools and one casks storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The wet pools range from forty years old to the most modern pool in the US Department of Energy (DOE) complex. The near-term objective is moving the fuel in the older wet storage facilities to

  7. MANAGING SPENT NUCLEAR FUEL WASTES AT THE IDAHO NATIONAL LABORATORY

    International Nuclear Information System (INIS)

    The Idaho National Engineering Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy is in part due to the history of the INL as the National Reactor Testing Station, in part to its mission to recover highly enriched uranium from SNF and in part to it's mission to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facility, some dating back 50 years in the site history. The success of the INL SNF program is measured by its ability to: (1) achieve safe existing storage, (2) continue to receive SNF from other locations, both foreign and domestic, (3) repackage SNF from wet storage to interim dry storage, and (4) prepare the SNF for dispositioning in a federal repository. Because of the diversity in the SNF and the facilities at the INL, the INL is addressing almost very condition that may exist in the SNF world. Many of solutions developed by the INL are applicable to other SNF storage sites as they develop their management strategy. The SNF being managed by the INL are in a variety of conditions, from intact assemblies to individual rods or plates to powders, rubble, and metallurgical mounts. Some of the fuel has been in wet storage for over forty years. The fuel is stored bare, or in metal cans and either wet under water or dry in vaults, caissons or casks. Inspections have shown varying degrees of corrosion and degradation of the fuel and the storage cans. Some of the fuel has been recanned under water, and the conditions of the fuel inside the second or third can are unknown. The fuel has been stored in one of 10 different facilities: five wet pools and one casks storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The wet pools range from forty years old to the most modern pool in the US Department of Energy (DOE) complex. The near-term objective is moving the fuel in the older wet storage facilities to

  8. Status of dry storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Spent-fuel storage has been identified as the key element of spent-fuel management. Over 45,000 t of spent water reactor fuel has been discharged worldwide, of which only ∼7% has been reprocessed. Estimates by the Organization for Economic Cooperation and Development and the International Atomic Energy Agency (IAEA) indicate that the amount of spent fuel being generated will increase significantly. About 200,000 t of heavy metal of spent fuel could be accumulated by the year 2000. Many countries are involved in the development of new ways, including dry storage, for handling and storing the spent fuel. These new technologies require new reviews and perhaps new approaches for domestic and international safeguards. The IAEA has been involved in surveying worldwide experience related to the storage of spent fuel. This paper summarizes the efforts of an international working group to survey the experience with dry storage and related innovations that might have an impact on safeguarding procedures in the future

  9. Preoperational Environmental Survey for the Spent Nuclear Fuel (SNF) Project Facilities

    Energy Technology Data Exchange (ETDEWEB)

    MITCHELL, R.M.

    2000-10-12

    This document represents the report for environmental sampling of soil, vegetation, litter, cryptograms, and small mammals at the Spent Nuclear Fuel Project facilities located in 100 K and 200 East Areas in support of the preoperational environmental survey.

  10. Preoperational Environmental Survey for the Spent Nuclear Fuel (SNF) Project Facilities

    Energy Technology Data Exchange (ETDEWEB)

    MITCHELL, R.M.

    2000-09-28

    This document represents the report for environmental sampling of soil, vegetation, litter, cryptograms, and small mammals at the Spent Nuclear Fuel Project facilities located in 100 K and 200 East Areas in support of the preoperational environmental survey.

  11. Review of the KBS II plan for handling and final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

    The Swedish utilities programme for disposal of spent nuclear fuel elements (KBS II) is summarized. Comments and criticism to the programme are given by experts from several foreign or international institutions. (L.E.)

  12. Investigations of possibilities to dispose of spent nuclear fuel in Lithuania: a model case. Summary Report

    International Nuclear Information System (INIS)

    The objective of this report is to demonstrate that in Lithuania there exist adequate competences in the area of direct disposal of spent nuclear fuel and that in principle it is possible to implement direct disposal of spent nuclear fuel in Lithuania in a safe way. The objective is not to indicate that direct disposal of spent nuclear fuel will take place in Lithuania. What option to use for the eventual disposal of the Lithuanian spent nuclear fuel is to a large extent a political decision, and this report will be an important input to such a decision once it will have to be taken. A summary volume contains a foreword, executive summary, introduction and conclusions

  13. FINAL REPORT (PART 1). RADIOLYTIC AND THERMAL PROCESSES RELEVANT TO DRY STORAGE OF SPENT NUCLEAR FUELS

    Science.gov (United States)

    The scientific and engineering demands of the Department of Energy (DOE) Environmental Restoration and Waste Management tasks are enormous. For example, several thousand metric tons of metallic uranium spent nuclear fuel (SNF) remain in water storage awaiting disposition. Of this...

  14. Preoperational Environmental Survey for the Spent Nuclear Fuel (SNF) Project Facilities

    International Nuclear Information System (INIS)

    This document represents the report for environmental sampling of soil, vegetation, litter, cryptograms, and small mammals at the Spent Nuclear Fuel Project facilities located in 100 K and 200 East Areas in support of the preoperational environmental survey

  15. Safety aspects of spent nuclear fuel shipment from 'Vinca' Institute to the Russian Federation

    International Nuclear Information System (INIS)

    The main safety aspects were analyzed and possible radiation impact on the personnel, population and environment during spent nuclear fuel preparation for shipment to the Russian Federation was estimated. Basic safety principals in elaborating repackaging technology are specified. (author)

  16. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    Science.gov (United States)

    Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

    2014-06-10

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  17. Analysis of storage conditions of the spent nuclear fuel on the dry storage platform

    International Nuclear Information System (INIS)

    The analysis of temperature measurements of air which cools storage containers of the spent nuclear fuel on the dry storage platform of the Zaporizhska NPP is carried out. Factors which influence a thermal condition of storage containers are defined, and their influence on results of temperature measurements is analyzed. Recommendations about improvement of system of the temperature control of storage containers of spent nuclear fuel are given.

  18. Deployment evaluation methodology for the electrometallurgical treatment of DOE-EM spent nuclear fuel

    International Nuclear Information System (INIS)

    Part of the Department of Energy (DOE) spent nuclear fuel (SNF) inventory may require some type of treatment to meet acceptance criteria at various disposition sites. The current focus for much of this spent nuclear fuel is the electrometallurgical treatment process under development at Argonne National Laboratory. Potential flowsheets for this treatment process are presented. Deployment of the process for the treatment of the spent nuclear fuel requires evaluation to determine the spent nuclear fuel program need for treatment and compatibility of the spent nuclear fuel with the process. The evaluation of need includes considerations of cost, technical feasibility, process material disposition, and schedule to treat a proposed fuel. A siting evaluation methodology has been developed to account for these variables. A work breakdown structure is proposed to gather life-cycle cost information to allow evaluation of alternative siting strategies on a similar basis. The evaluation methodology, while created specifically for the electrometallurgical evaluation, has been written such that it could be applied to any potential treatment process that is a disposition option for spent nuclear fuel. Future work to complete the evaluation of the process for electrometallurgical treatment is discussed

  19. Government/industry development of a spent nuclear fuel dry transfer system

    International Nuclear Information System (INIS)

    The United States Department of Energy (DOE) is currently engaged in a cooperative program with the Electric Power Research Institute (EPRI) to develop a spent nuclear fuel dry transfer system (DTS). The system will enable the transfer of individual spent nuclear fuel assemblies between a conventional top loading cask and multi-purpose canister in a shielded overpack, or accommodate spent nuclear fuel transfers between two conventional casks. The DTS has several significant applications and could benefit the Federal waste management system and utilities in number of ways. It has the potential to: permit shutdown reactor sites to decommission pools; provide capability at interim storage facilities to transfer assemblies from small transportation casks to sealed canisters; provide capability at reactor sites with limited crane capacity to transfer assemblies into large packages to facilitate on-site storage in larger capacity casks; allow recovery operations at shutdown reactor sites with independent spent nuclear fuel storage installations; provide a means for utilities that can presently handle only a truck cask to utilize a rail cask; allow transfers of spent nuclear fuel from existing utility dual purpose systems into alternative systems if required, without returning to the reactor storage pool; and support existing and future DOE spent nuclear fuel management activities. The project is managed by a Technical Management Committee consisting of DOE and EPRI representatives and includes a member from Battelle Pacific Northwest National Laboratory. (authors)

  20. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-12

    The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in the emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Multiple boiling water reactor (BWR) and pressurized water reactor (PWR) disposal container designs are needed to accommodate the expected range of spent fuel assemblies and provide long-term confinement of the commercial SNF. The disposal container will include outer and inner cylinder walls, outer cylinder lids (two on the top, one on the bottom), inner cylinder lids (one on the top, one on the bottom), and an internal metallic basket structure. Exterior labels will provide a means by which to identify the disposal container and its contents. The two metal cylinders, in combination with the cladding, Emplacement Drift System, drip shield, and natural barrier, will support the design philosophy of defense-in-depth. The use of materials with different

  1. An integrated methodology to evaluate a spent nuclear fuel storage system

    International Nuclear Information System (INIS)

    This study introduced a methodology that can be applied for development of a dry storage system for spent nuclear fuels. It consisted of several design activities that includes development of a simplified program to analyze the amount of spent nuclear fuels from reflecting the practical situation in spent nuclear fuel management and a simplified program to evaluate the cost of 4 types of representing storage system to choose the most competitive option considering economic factor. As verification of the implementation of the reference module to practical purpose, a simplified thermal analysis code was suggested that can see fulfillment of limitation of temperature in long term storage and oxidation analysis. From the thermal related results, the reference module can accommodate full range of PHWR spent nuclear fuels and significant portion of PWR ones too. From the results, the reference storage system can be concluded that has fulfilled the important requirements in terms of long term integrity and radiological safety. Also for the purpose of solving scattered radiation along with deep penetration problems in cooling storage system, small but efficient design alternation was suggested together with its efficiency that can reduce scattered radiation by 1/3 from the original design. Along with the countermeasure for the shielding problem, in consideration of PWR spent nuclear fuels, simplified criticality analysis methodology retaining conservativeness was proposed. The results show the reference module is efficient low enrichment PWR spent nuclear fuel and even relatively high enrichment fuels too if burnup credit is taken. As conclusive remark, the methodology is simple but efficient to plan a concept design of convective cooling type of spent nuclear fuels storage. It can be also concluded that the methodology derived in this study and the reference module has feasibility in practical implementation to mitigate the current complex situation in spent fuel

  2. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

  3. Spent nuclear fuel project high-level information management plan

    Energy Technology Data Exchange (ETDEWEB)

    Main, G.C.

    1996-09-13

    This document presents the results of the Spent Nuclear Fuel Project (SNFP) Information Management Planning Project (IMPP), a short-term project that identified information management (IM) issues and opportunities within the SNFP and outlined a high-level plan to address them. This high-level plan for the SNMFP IM focuses on specific examples from within the SNFP. The plan`s recommendations can be characterized in several ways. Some recommendations address specific challenges that the SNFP faces. Others form the basis for making smooth transitions in several important IM areas. Still others identify areas where further study and planning are indicated. The team`s knowledge of developments in the IM industry and at the Hanford Site were crucial in deciding where to recommend that the SNFP act and where they should wait for Site plans to be made. Because of the fast pace of the SNFP and demands on SNFP staff, input and interaction were primarily between the IMPP team and members of the SNFP Information Management Steering Committee (IMSC). Key input to the IMPP came from a workshop where IMSC members and their delegates developed a set of draft IM principles. These principles, described in Section 2, became the foundation for the recommendations found in the transition plan outlined in Section 5. Availability of SNFP staff was limited, so project documents were used as a basis for much of the work. The team, realizing that the status of the project and the environment are continually changing, tried to keep abreast of major developments since those documents were generated. To the extent possible, the information contained in this document is current as of the end of fiscal year (FY) 1995. Programs and organizations on the Hanford Site as a whole are trying to maximize their return on IM investments. They are coordinating IM activities and trying to leverage existing capabilities. However, the SNFP cannot just rely on Sitewide activities to meet its IM requirements

  4. MTR radiological database for SRS spent nuclear fuel facilities

    International Nuclear Information System (INIS)

    A database for radiological characterization of incoming Material Test Reactor (MTR) fuel has been developed for application to the Receiving Basin for Offsite Fuels (RBOF) and L-Basin spent fuel storage facilities at the Savannah River Site (SRS). This database provides a quick quantitative check to determine if SRS bound spent fuel is radiologically bounded by the Reference Fuel Assembly used in the L-Basin and RBOF authorization bases. The developed database considers pertinent characteristics of domestic and foreign research reactor fuel including exposure, fuel enrichment, irradiation time, cooling time, and fuel-to-moderator ratio. The supplied tables replace the time-consuming studies associated with authorization of SRS bound spent fuel with simple hand calculations. Additionally, the comprehensive database provides the means to overcome resource limitations, since a series of simple, yet conservative, hand calculations can now be performed in a timely manner and replace computational and technical staff requirements

  5. Handling of final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

    In this report the various facilities incorporated in the proposed handling chain for spent fuel from the power stations to the final repository are discribed. Thus the geological conditions which are essential for a final repository is discussed as well as the buffer and canister materials and how they contribute towards a long-term isolation of the spent fuel. Furthermore one chapter deals with leaching of the deposited fuel in the event that the canister is penetrated as well as the transport mechanisms which determine the migration of the radioactive substances through the buffer material. The dispersal processes in the geosphere and the biosphere are also described together with the transfer mechanisms to the ecological systems as well as radiation doses. Finally a summary is given of the safety analysis of the proposed method for the handling and final storage of the spent fuel. (E.R.)

  6. State fund of decommissioning of nuclear installations and handling of spent nuclear fuels and nuclear wastes (Slovak Republic)

    International Nuclear Information System (INIS)

    State Fund for Decommissioning of Nuclear Installations and Handling of Spent Nuclear Fuels and Nuclear Wastes was established by the Act 254/1994 of the National Council of the Slovak Republic as a special-purpose fund which concentrates financial resources intended for decommissioning of nuclear installations and for handling of spent nuclear fuels and radioactive wastes. The Act was amended in 2000, 2001 and 2002. The Fund is legal entity and independent from operator of nuclear installations Slovak Power Facilities Inc. The Fund is headed by Director, who is appointed and recalled by Minister of Economy of the Slovak Republic. Sources of the Fund are generated from: a) contributions by nuclear installation operators; b) penalties imposed by Nuclear Regulatory Authority of the Slovak Republic upon natural persons and legal entities pursuant to separate regulation; c) bank credits; d) interest on Fund deposits in banks; e) grants from State Budget; f) other sources as provided by special regulation. Fund resources may be used for the following purposes: a) decommissioning of nuclear installations; b) handling of spent nuclear fuels and radioactive wastes after the termination of nuclear installation operation; c) handling of radioactive wastes whose originator is not known, including occasionally seized radioactive wastes and radioactive materials stemming from criminal activities whose originator is not known, as confirmed by Police Corps investigator or Ministry of Health of the Slovak Republic; d) purchase of land for the establishment of nuclear fuel and nuclear waste repositories; e) research and development in the areas of decommissioning of nuclear installations and handling of nuclear fuels and radioactive wastes after the termination of the operation of nuclear installations; f) selection of localities, geological survey, preparation, design, construction, commissioning, operation and closure of repositories of spent nuclear fuels and radioactive wastes

  7. SACSESS – the EURATOM FP7 project on actinide separation from spent nuclear fuels

    OpenAIRE

    Bourg Stéphane; Geist Andreas; Narbutt Jerzy

    2015-01-01

    Recycling of actinides by their separation from spent nuclear fuel, followed by transmutation in fast neutron reactors of Generation IV, is considered the most promising strategy for nuclear waste management. Closing the fuel cycle and burning long-lived actinides allows optimizing the use of natural resources and minimizing the long-term hazard of high-level nuclear waste. Moreover, improving the safety and sustainability of nuclear power worldwide. This paper presents the activities strivin...

  8. Use of axial burnup distribution profile in the nuclear safety analysis of spent nuclear fuel storage for WWER reactors in Ukraine

    International Nuclear Information System (INIS)

    The nuclear safety analysis of spent fuel storages taking into account fuel burnup should allow for burnup distribution along the height of the assembly. We propose a method based on an analysis of the axial burnup profiles of spent fuel assemblies. This method can be used in nuclear safety justification of spent fuel management and storage systems

  9. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    2005-01-01

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based stor

  10. Evaluation of Radiation Impacts of Spent Nuclear Fuel Storage (SNFS-2) of Chernobyl NPP - 13495

    International Nuclear Information System (INIS)

    Radiation effects are estimated for the operation of a new dry storage facility for spent nuclear fuel (SNFS-2) of Chernobyl NPP RBMK reactors. It is shown that radiation exposure during normal operation, design and beyond design basis accidents are minor and meet the criteria for safe use of radiation and nuclear facilities in Ukraine. (authors)

  11. Analysis of Spent Nuclear Fuel Imaging Using Multiple Coulomb Scattering of Cosmic Muons

    CERN Document Server

    Chatzidakis, Stylianos; Tsoukalas, Lefteri H

    2016-01-01

    Cosmic ray muons passing through matter lose energy from inelastic collisions with electrons and are deflected from nuclei due to multiple Coulomb scattering. The strong dependence of scattering on atomic number Z and the recent developments on position sensitive muon detectors indicate that multiple Coulomb scattering could be an excellent candidate for spent nuclear fuel imaging. Muons present significant advantages over existing monitoring and imaging techniques and can play a central role in monitoring nuclear waste and spent nuclear fuel stored in dense well shielded containers. The main purpose of this paper is to investigate the applicability of multiple Coulomb scattering for imaging of spent nuclear fuel dry casks stored within vertical and horizontal commercial storage dry casks. Calculations of muon scattering were performed for various scenarios, including vertical and horizontal fully loaded dry casks, half loaded dry casks, dry casks with one row of fuel assemblies missing, dry casks with one fu...

  12. Site selection - siting of the final repository for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    2011-03-15

    SKB has selected Forsmark as the site for the final repository for spent nuclear fuel. The site selection is the end result of an extensive siting process that began in the early 1990s. The strategy and plan for the work was based on experience from investigations and development work over a period of more than ten years prior to then. This document describes the siting work and SKB's choice of site for the final repository. It also presents the information on which the choice was based and the reasons for the decisions made along the way. The document comprises Appendix PV to applications under the Nuclear Activities Act and the Environmental Code for licences to build and operate an encapsulation plant adjacent to the central interim storage facility for spent nuclear fuel in Oskarshamn, and to build and operate a final repository for spent nuclear fuel in Forsmark in Oesthammar Municipality

  13. RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA

    Energy Technology Data Exchange (ETDEWEB)

    Kelly Cummins; Igor Bolshinsky; Ken Allen; Tihomir Apostolov; Ivaylo Dimitrov

    2009-07-01

    In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required to complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.

  14. Site selection - siting of the final repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    SKB has selected Forsmark as the site for the final repository for spent nuclear fuel. The site selection is the end result of an extensive siting process that began in the early 1990s. The strategy and plan for the work was based on experience from investigations and development work over a period of more than ten years prior to then. This document describes the siting work and SKB's choice of site for the final repository. It also presents the information on which the choice was based and the reasons for the decisions made along the way. The document comprises Appendix PV to applications under the Nuclear Activities Act and the Environmental Code for licences to build and operate an encapsulation plant adjacent to the central interim storage facility for spent nuclear fuel in Oskarshamn, and to build and operate a final repository for spent nuclear fuel in Forsmark in Oesthammar Municipality

  15. Development of Integrated Analyzing and Training Simulator for Spent Nuclear Fuel Pool, CAREPooL

    International Nuclear Information System (INIS)

    The Comprehensive Analyzer of Real Estimation for spent fuel POOL (CAREPOOL) has been developed for evaluating temperature and criticality of a spent nuclear fuel pool (SFP) during the normal and accident conditions. The CAREPOOL provides four main functions- management of spent nuclear fuel, decay heat Transactions of the Korean Nuclear Society Spring Meeting Jeju, Korea, May 7-8, 2015 calculation by ORIGEN-S code, estimation of the time to boil/fuel uncovering by thermal-hydraulics calculations, criticality evaluation by Helios code. All of these are integrated into the GUI based CAREPOOL system. The CAREPOOL would be very beneficial to nuclear power plant operator and trainee who have responsibility for the SFP operation

  16. A Study On The Initial Characteristics Of Domestic Spent Nuclear Fuels For Long Term DRY Storage

    International Nuclear Information System (INIS)

    During the last three decades, South Korean nuclear power plants have discharged about 5,950 tons of spent fuel and the maximum burn-up reached 55 GWd/MTU in 2002. This study was performed to support the development of Korean dry spent fuel storage alternatives. First, we chose V5H-17 Χ 17 and KSFA-16 Χ 16 as representative domestic spent fuels, considering current accumulation and the future generation of the spent fuels. Examination reveals that their average burn-ups have already increased from 33 to 51 GWd/MTU and from 34.8 to 48.5 GWd/MTU, respectively. Evaluation of the fuel characteristics shows that at the average burn-up of 42 GWd/MTU, the oxide thickness, hydrogen content, and hoop stress ranged from 30 ∼ 60 μm, 250 ∼ 500 ppm, and 50 ∼ 75 MPa, respectively. But when burn-up exceeds 55 GWd/MTU, those characteristics can increase up to 100 μm, 800 ppm, and 120 MPa, respectively, depending on the power history. These results demonstrate that most Korean spent nuclear fuels are expected to remain within safe bounds during long-term dry storage, however, the excessive hoop stress and hydrogen concentration may trigger the degradation of the spent fuel integrity early during the long-term dry storage in the case of high burn-up spent fuels exceeding 45 GWd/MTU

  17. Conceptual design report for the ICPP spent nuclear fuel dry storage project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The conceptual design is presented for a facility to transfer spent nuclear fuel from shipping casks to dry storage containers, and to safely store those containers at ICPP at INEL. The spent fuels to be handled at the new facility are identified and overall design and operating criteria established. Physical configuration of the facility and the systems used to handle the SNF are described. Detailed cost estimate for design and construction of the facility is presented.

  18. Near-field chemistry of the spent nuclear fuel repository; Kemialliset vuorovaikutukset kaeytetyn ydinpolttoaineen loppusijoitustilan laehialueella

    Energy Technology Data Exchange (ETDEWEB)

    Kumpulainen, H.; Lehikoinen, J.; Muurinen, A.; Ollila, K. [VTT Chemical Technology, Espoo (Finland). Industrial Physics

    1998-07-01

    Factors affecting near-field chemistry of the spent nuclear fuel repository as well as the involved mutual interactions are described on the basis of literature. The most important processes in the near-field (spent-fuel, canister and bentonite) are presented. The related examples on near-field chemistry models shed light on the extensive problematics of near-field chemistry. (authors) 80 refs.

  19. International conference on management of spent fuel from nuclear power reactors. Book of extended synopses

    International Nuclear Information System (INIS)

    This document contains 48 extended synopses of the International Conference on Management of Spent Fuel from Nuclear Power Reactors. The major topics covered related to national programmes in spent fuel management as well as regional trends, technology and safety/security aspects of wet and dry storage, licensing and regulation, quality assurance, design control, operating experience, R and D, and special aspects of spent fuel storage including in-service inspection, robotics, heat removal, and other engineering considerations. Each of the extended synopses was indexed separately

  20. Hot isostatic pressing of ceramic waste from spent nuclear fuel

    International Nuclear Information System (INIS)

    Argonne National Laboratory has developed a process to immobilize waste salt containing fission products, uranium, and transuranic elements as chlorides in a glass-bonded ceramic waste form. This salt was generated in the electrorefining operation used in electrometallurgical treatment of spent Experimental Breeder Reactor-II fuel. The ceramic waste process culminated with a hot isostatic pressing operation. This paper reviews the installation and operation of a hot isostatic press in a radioactive environment. Processing conditions for the hot isostatic press are presented for non-irradiated material and irradiated material. Sufficient testing was performed to demonstrate that a hot isostatic press could be used as the final step of the processing of ceramic waste for the electrometallurgical spent fuel treatment process

  1. Risk of transporting spent nuclear fuel by truck

    International Nuclear Information System (INIS)

    The risk methodology used to evaluate the risk in shipping spent fuel includes: 1) a description of the spent fuel transport system, 2) identification of potential release sequences, 3) evaluation of the probabilities and consequences of the releases, and 4) calculation and assessment of the risk. The system description includes projected industry characteristics, amounts to be shipped, shipping package descriptions, material characteristics, transport mode, transport routes used and weather and population distribution information. Release sequences are identified by fault tree analysis tehniques. Releases are evaluated using package failure data, normal transport and transport accident environment data and mathematical models for material dispersion and resultant health effects. This information is combined to calculate the shipping system risk which is compared to other known risks. The data may be further analyzed to determine the primary contributors to the risk and identify possible methods for reducing the risk, if the current risk level is judged by society to be unacceptable

  2. Safety analysis of disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    The spent fuel from the Olkiluoto NPP (TVO I and II) is planned to be disposed of in a repository to be constructed at a depth of about 500 meters in the crystalline bedrock. The thesis is dealing with the safety analysis of the disposal. The main topics presented in the thesis are: (1) The amount of radioactive properties of the spent fuel, (2) The canister design and the planned disposal concept, (3) The results of the preliminary site investigations, (4) Discussion of the multi-barrier principle, (5) The general principles and methodology of the TVO-92 safety analysis, (6) Groundwater flow analysis, (7) Durability and behaviour of the canister, (8) Biosphere analysis and reference scenario, and (9) The sensitivity and uncertainty analyses. (246 refs., 75 figs., 44 tabs.)

  3. 75 FR 45167 - Notice of Public Workshop on a Potential Rulemaking for Spent Nuclear Fuel Reprocessing Facilities

    Science.gov (United States)

    2010-08-02

    ... civilian nuclear power globally and close the nuclear fuel cycle through reprocessing spent fuel and... Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor... regulations in 10 CFR Part 171, ``Annual Fees for Reactor Licenses and Fuel Cycle Licenses and......

  4. Intermediate storage and some issues concerning the final storage of spent nuclear fuel at Cernavoda NPP

    International Nuclear Information System (INIS)

    A major objective of 'Nuclearelectrica' National Society (SNN) is to implement nuclear power projects with respect to the sustainable development concept. Examples are given regarding spent fuel and radioactive waste management. Managing the spent fuel according with the concept of sustainable development is demonstrated by the SNN's policy and target objectives of the associated strategy. Spent fuel management strategy is implemented through an optimized staged process. One of the most important stages was completed by commissioning the first unit of the Dry Interim Spent Fuel Storage Facility at Cernavoda NPP site. The dry storage facility is based on the MACSTOR technology , the best proved design of Atomic Energy of Canada Limited and consists of concrete modules, each of them having a 12,000 spent fuel bundles capacity. The designed lifetime of the facility is, at least, 50 years and the present site can accommodate the spent fuel generated by two units during 30 years of operation. Spent fuel storage is the last and the most important stage of disposal. In Romania, by the Government's Ordinance 11/2003 approved by the Law 320/2003 there were assigned the responsibilities within the national waste management system. The newly established National Agency for Radioactive Waste Management will have a fundamental task, being responsible for the end stage of fuel cycle by selecting for the spent fuel disposal an appropriate repository site accepted by the public and a safe and efficient technology with a reasonable acceptable environmental impact. (author)

  5. OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, Ian C [ORNL; Sly, Nicholas C [ORNL; Michel-Sendis, Franco [OECD Nuclear Energy Agency

    2014-01-01

    Experimental data on the isotopic concentrations in irradiated nuclear fuel represent one of the primary methods for validating computational methods and nuclear data used for reactor and spent fuel depletion simulations that support nuclear fuel cycle safety and safeguards programs. Measurement data have previously not been available to users in a centralized or searchable format, and the majority of accessible information has been, for the most part, limited to light-water-reactor designs. This paper describes a recent initiative to compile spent fuel benchmark data for additional reactor designs used throughout the world that can be used to validate computer model simulations that support nuclear energy and nuclear safeguards missions. Experimental benchmark data have been expanded to include VVER-440, VVER-1000, RBMK, graphite moderated MAGNOX, gas cooled AGR, and several heavy-water moderated CANDU reactor designs. Additional experimental data for pressurized light water and boiling water reactor fuels has also been compiled for modern assembly designs and more extensive isotopic measurements. These data are being compiled and uploaded to a recently revised structured and searchable database, SFCOMPO, to provide the nuclear analysis community with a centrally-accessible resource of spent fuel compositions that can be used to benchmark computer codes, models, and nuclear data. The current version of SFCOMPO contains data for eight reactor designs, 20 fuel assembly designs, more than 550 spent fuel samples, and measured isotopic data for about 80 nuclides.

  6. Egyptian Proposed Strategy of Spent Fuel Management from Nuclear Power Reactors

    International Nuclear Information System (INIS)

    This paper presents a proposed policy and strategy of the Arab Republic of Egypt towards the management of spent fuel from the nuclear power reactors. The proposed Egyptian strategy supports the free international trade of nuclear materials and services and adheres to Non Proliferation Treaty (NPT) and other institutional frameworks aimed at promoting the peaceful use of nuclear power in all countries. Considering the spent fuel is a main part of nuclear fuel cycle, therefore the Nuclear Power Plants Authority (NPPA) is responsible for the assurance, safety, secured and stable supply of all services and materials of nuclear fuel cycle within a long term contracts. Egypt has taken the decision to adopt an open fuel cycle for the first nuclear power plant, i.e. no reprocessing of spent fuel. NPPA would develop at an early stage a conceptual plan describing all important steps leading to the final disposal of spent fuel and radioactive waste in Egypt utilizing fully the national and international experience and the capabilities of international cooperation. (author)

  7. Track 9: fuel cycle, spent fuel, decommissioning, and waste management. Dry storage of commercial spent nuclear fuel. Panel Discussion

    International Nuclear Information System (INIS)

    Full text of publication follows: Currently, dry storage of spent nuclear fuel is a mature technology with a firm technical basis when the burnup of fuel is 45 GWd/tonne U for up to 100-yr duration. A technical basis for the safe extension of dry storage to the higher burnups and longer times needs to be established. In particular, the expected behavior of cask and fuel materials must be established. Experts from the U.S. Nuclear Regulatory Commission, the American Society for Testing and Materials, utilities storing fuel, the Electric Power Research Institute, and national laboratories will discuss the current status of that database, what data need to be obtained, what programs are in place to obtain data to augment the basis, and the results of those programs. (authors)

  8. AIR SHIPMENT OF HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL FROM ROMANIA AND LIBYA

    Energy Technology Data Exchange (ETDEWEB)

    Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

    2010-07-01

    In June 2009 Romania successfully completed the world’s first air shipment of highly enriched uranium (HEU) spent nuclear fuel transported in Type B(U) casks under existing international laws and without special exceptions for the air transport licenses. Special 20-foot ISO shipping containers and cask tiedown supports were designed to transport Russian TUK 19 shipping casks for the Romanian air shipment and the equipment was certified for all modes of transport, including road, rail, water, and air. In December 2009 Libya successfully used this same equipment for a second air shipment of HEU spent nuclear fuel. Both spent fuel shipments were transported by truck from the originating nuclear facilities to nearby commercial airports, were flown by commercial cargo aircraft to a commercial airport in Yekaterinburg, Russia, and then transported by truck to their final destinations at the Production Association Mayak facility in Chelyabinsk, Russia. Both air shipments were performed under the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI). The Romania air shipment of 23.7 kg of HEU spent fuel from the VVR S research reactor was the last of three HEU fresh and spent fuel shipments under RRRFR that resulted in Romania becoming the 3rd RRRFR participating country to remove all HEU. Libya had previously completed two RRRFR shipments of HEU fresh fuel so the 5.2 kg of HEU spent fuel air shipped from the IRT 1 research reactor in December made Libya the 4th RRRFR participating country to remove all HEU. This paper describes the equipment, preparations, and license approvals required to safely and securely complete these two air shipments of spent nuclear fuel.

  9. Fresh and Spent Nuclear Fuel Repatriation from the IRT-2000 Research Reactor Facility, Sofia, Bulgaria

    Energy Technology Data Exchange (ETDEWEB)

    K. J. Allen; T. G. Apostolov; I. S. Dimitrov

    2009-03-01

    The IRT 2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned.

  10. Investigation of Thermal Processes at Dry Storage of Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    For good safety dry storage of the spent nuclear fuel of the Zaporizhska nuclear power plant (Ukraine) investigations of thermal processes are carried out. Researches were carried out by the solution of the conjugate problems of heat exchange. The free convection of ventilating air and helium into basket of storage, and radiative heat exchange in the container was considered. As a result temperature fields of the container and air in windless day and temperature fields of the spent fuel assemblies with identical and various energy-release inside the container of a storage are obtained. Results of investigations can be used for construction of safety nearstations dry storages in other atomic power plants of Ukraine or at open dry storages of the spent nuclear fuel in other countries. (author)

  11. Equivalent thermal conductivity of the storage basket with spent nuclear fuel of VVER-1000 reactors

    International Nuclear Information System (INIS)

    Due to limitation of computation resources and/or computation time many thermal problems require to use simplified geometrical models with equivalent thermal properties. A new method for definition of equivalent thermal conductivity of spent nuclear fuel storage casks is proposed. It is based on solving the inverse heat conduction problem. For the proposed method two approaches for equivalent thermal conductivity definition were considered. In the first approach a simplified model in conjugate formulation is used, in the second approach a simplified model of solid body which allows an analytical solution is used. For safety ensuring during all time of spent nuclear fuel storage the equivalent thermal conductivity was calculated for different storage years. The calculated equivalent thermal conductivities can be used in thermal researches for dry spent nuclear fuel storage safety.

  12. Evolution of spent nuclear fuel in dry storage conditions for millennia and beyond

    International Nuclear Information System (INIS)

    Significant amounts of spent uranium dioxide nuclear fuel are accumulating worldwide from decades of commercial nuclear power production. While such spent fuel is intended to be reprocessed or disposed in geologic repositories, out-of-reactor radiation damage from alpha decay can be detrimental to its structural stability. Here we report on an experimental study in which radiation damage in plutonium dioxide, uranium dioxide samples doped with short-lived alpha-emitters and urano-thorianite minerals have been characterized by XRD, transmission electron microscopy, thermal desorption spectrometry and hardness measurements to assess the long-term stability of spent nuclear fuel to substantial alpha-decay doses. Defect accumulation is predicted to result in swelling of the atomic structure and decrease in fracture toughness; whereas, the accumulation of helium will produce bubbles that result in much larger gaseous-induced swelling that substantially increases the stresses in the constrained spent fuel. Based on these results, the radiation-ageing of highly-aged spent nuclear fuel over more than 10,000 years is predicted

  13. Spent Nuclear Fuel Cask and Storage Monitoring with 4He Scintillation Fast Neutron Detectors

    International Nuclear Information System (INIS)

    With this increasing quantity of spent nuclear fuel being stored at nuclear plants across S. Korea, the demand exists for building a long-term disposal facility. However, the Korean government first requires a detailed plan for the monitoring and certification of spent fuel. Several techniques have been developed and applied for the purpose of spent fuel monitoring, including the digital Cerenkov viewing device (DCVD), spent fuel attribute tester (SFAT), and FORK detector. Conventional gamma measurement methods, however, suffer from a lack of nuclear data and interfering background radiation. To date, the primary method of neutron detection for spent fuel monitoring has been through the use of thermal neutron detectors such as 3He and BF3 proportional counters. Unfolding the neutron spectrum becomes extremely complicated. In an attempt to overcome these difficulties, a new fast neutron measurement system is currently being developed at the University of Florida. This system is based on the 4He scintillation detector invented by Arktis Radiation Detectors Ltd. These detectors are a relatively new technological development and take advantage of the high 4He cross-section for elastic scattering at fast neutron energies, particularly the resonance around 1 MeV. This novel 4He scintillation neutron detector is characterized by its low electron density, leading to excellent gamma rejection. This detector also has a fast response time on the order of nanoseconds and most importantly, preserves some neutron energy information since no moderator is required. Additionally, these detectors rely on naturally abundant 4He as the fill gas. This study proposes a new technique using the neutron spectroscopy features of 4He scintillation detectors to maintain accountability of spent fuel in storage. This research will support spent fuel safeguards and the detection of fissile material, in order to minimize the risk of nuclear proliferation and terrorism. This study conducted a

  14. Thermal analysis of cold vacuum drying of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Piepho, M.G.

    1998-07-20

    The thermal analysis examined transient thermal and chemical behavior of the Multi canister Overpack (MCO) container for a broad range of cases that represent the Cold Vacuum Drying (CVD) processes. The cases were defined to consider both normal and off-normal operations at the CVD Facility for an MCO with Mark IV N, Reactor spent fuel in four fuel baskets and one scrap basket. This analysis provides the basis for the MCO thermal behavior at the CVD Facility for its Phase 2 Safety Analysis Report (revision 4).

  15. Spent nuclear fuel and high-level radioactive waste handling, storage and disposal in the Slovak Republic

    International Nuclear Information System (INIS)

    The following topics are discussed in detail: Concept of the ultimate stages of spent fuel management in the Slovak Republic; Spent fuel and radioactive waste management strategy; and Spent nuclear fuel interim storage facility. The legal framework and government decisions are described. Slovakia’s policy is in line with the 1988 Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management. The strategy for the future is such that spent nuclear fuel should be reprocessed and radioactive waste should be disposed of in a deep geological repository. (orig.)

  16. Projections of spent fuel to be discharged by the U.S. nuclear power industry

    International Nuclear Information System (INIS)

    Calculated properties of spent fuel projected to be discharged and accumulated by the U.S. nuclear power industry through the year 2031 A.D. are presented. The projections are based on installed nuclear capacities of 380 and 543 GW(e) in the year 2000 and 2030, respectively. They include compilations of the grams of the elements, curies of radioactivity, thermal decay power, photon and neutron emission rates, and radiotoxicities of the assemblies that are accumulated at a Spent Unreprocessed Fuel Facility (SURFF), allowing for delays of 5 and 10 years before shipment to SURFF

  17. Total Spent Nuclear Fuel Back-End Costs For Ignalina NPP

    International Nuclear Information System (INIS)

    Discussions are going on different acceptable scenarios for Ignalina NPP decommissioning. To choose the optimum alternative it is necessary to know the spent nuclear fuel back-end costs for Ignalina NPP. This paper deals with an analysis of such results for different decommissioning scenarios. The costs are estimated for the main three components: NPP decommissioning and short-lived waste treatment costs, spent nuclear fuel (SNF) and long-lived waste (LLW) storage costs, SNF and LLW disposal costs. Also, the annual flow of these costs is shown The estimated data are compared with the available data on Ignalina NPP and Sweden. (author)

  18. Chemical Reactivity Testing for the National Spent Nuclear Fuel Program. Quality Assurance Project Plan

    International Nuclear Information System (INIS)

    This quality assurance project plan (QAPjP) summarizes requirements used by Lockheed Martin Energy Systems, Incorporated (LMES) Development Division at Y-12 for conducting chemical reactivity testing of Department of Energy (DOE) owned spent nuclear fuel, sponsored by the National Spent Nuclear Fuel Program (NSNFP). The requirements are based on the NSNFP Statement of Work PRO-007 (Statement of Work for Laboratory Determination of Uranium Hydride Oxidation Reaction Kinetics.) This QAPjP will utilize the quality assurance program at Y-12, QA-101PD, revision 1, and existing implementing procedures for the most part in meeting the NSNFP Statement of Work PRO-007 requirements, exceptions will be noted

  19. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 2, Calculated activity profiles of spent nuclear fuel assembly hardware for pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Short, S.M.; Luksic, A.T.; Lotz, T.L.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report present a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from Laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  20. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 3, Calculated activity profiles of spent nuclear fuel assembly hardware for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Short, S.M.; Luksic, A.T.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly that is also radioactive and required disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volume 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  1. Thermal hydraulic and neutronic analysis of dry cask storage systems for spent nuclear fuels

    International Nuclear Information System (INIS)

    Interim spent fuel storage systems must provide for the safe receipt, handling, retrieval and storage of spent nuclear fuel before reprocessing or disposal. In the context of achieving these objectives, the following features of the design were taken into consideration for metal shielded type storage systems; to maintain fuel subcritical, to remove spent fuel residual heat, to provide for radiation protection. These features in the design of a dry cask storage system were analyzed by employing COBRA-SFS and SCALE4.4 (ORIGEN, XSDOSE, CSAS6 ) codes for normal operation of the system under study. In accordance with safety assurance limits of International Atomic Energy Authority (IAEA), appropriate designs for Dry Cask Storage Systems (DCSS) were reached for 33000, 45000, and 55000 MWd/t burnup values and 5 and 10 years of cooling periods for spent fuel to be stored (Table 1)

  2. U.S. spent nuclear fuel acceptance policy 2006 and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Edlow, J.; Brown, K. R. [Edlow International Company, Washington D.C., 20009 (United States)

    2002-07-01

    In 1992, a large group of research reactor operators joined with Edlow International Company (EIC) forming the Edlow Group and promote the interests of members before the U.S. government and to enhance the negotiating leverage of each member. The goal was to convince the U.S. Department of Energy (DOE) to resume acceptance of foreign research reactor spent fuel after a three-year suspension. Approximately four years later in May of 1996, largely due to the efforts of the Edlow Group, DOE, in consultation with the Department of State (DOS), issued a new research reactor spent nuclear fuel acceptance policy. After five years of successful spent fuel shipments, new issues have surfaced which led of the Edlow Group to represent the interests of the foreign research reactor community before the DOE. This paper discusses the new spent research reactor fuel issues, and proposes utilizing a reactivated Edlow Group to address these concerns. (author)

  3. U.S. spent nuclear fuel acceptance policy 2006 and beyond

    International Nuclear Information System (INIS)

    In 1992, a large group of research reactor operators joined with Edlow International Company (EIC) forming the Edlow Group and promote the interests of members before the U.S. government and to enhance the negotiating leverage of each member. The goal was to convince the U.S. Department of Energy (DOE) to resume acceptance of foreign research reactor spent fuel after a three-year suspension. Approximately four years later in May of 1996, largely due to the efforts of the Edlow Group, DOE, in consultation with the Department of State (DOS), issued a new research reactor spent nuclear fuel acceptance policy. After five years of successful spent fuel shipments, new issues have surfaced which led of the Edlow Group to represent the interests of the foreign research reactor community before the DOE. This paper discusses the new spent research reactor fuel issues, and proposes utilizing a reactivated Edlow Group to address these concerns. (author)

  4. Standard guide for characterization of spent nuclear fuel in support of geologic repository disposal

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This guide provides guidance for the types and extent of testing that would be involved in characterizing the physical and chemical nature of spent nuclear fuel (SNF) in support of its interim storage, transport, and disposal in a geologic repository. This guide applies primarily to commercial light water reactor (LWR) spent fuel and spent fuel from weapons production, although the individual tests/analyses may be used as applicable to other spent fuels such as those from research and test reactors. The testing is designed to provide information that supports the design, safety analysis, and performance assessment of a geologic repository for the ultimate disposal of the SNF. 1.2 The testing described includes characterization of such physical attributes as physical appearance, weight, density, shape/geometry, degree, and type of SNF cladding damage. The testing described also includes the measurement/examination of such chemical attributes as radionuclide content, microstructure, and corrosion product c...

  5. German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

    Directory of Open Access Journals (Sweden)

    A. Schwenk-Ferrero

    2013-01-01

    Full Text Available Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104 to 106 years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.

  6. Spent nuclear fuel discharges from US reactors 1992

    Energy Technology Data Exchange (ETDEWEB)

    1994-05-05

    This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactive Waste Management.

  7. The State of the WWER Nuclear Spent Fuel Management in Ukraine and Trends on the Optimal Choice of Spent Fuel Management Strategy

    International Nuclear Information System (INIS)

    The paper gives an overview of the operational state of the away from reactor interim spent fuel storage facility at the Zaporizhzhya NPP site, commissioned according to the dry storage cask technology that is developed by the US company Sierra Nuclear Corporation. The trends of the SFM at the WWER nuclear units were noted, which are caused by the advanced nuclear fuel implementation in order to improve the nuclear fuel utilization. (author)

  8. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix C, Savannah River Site Spent Nuclear Fuel Mangement Program

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The US Department of Energy (DOE) is engaged in two related decision making processes concerning: (1) the transportation, receipt, processing, and storage of spent nuclear fuel (SNF) at the DOE Idaho National Engineering Laboratory (INEL) which will focus on the next 10 years; and (2) programmatic decisions on future spent nuclear fuel management which will emphasize the next 40 years. DOE is analyzing the environmental consequences of these spent nuclear fuel management actions in this two-volume Environmental Impact Statement (EIS). Volume 1 supports broad programmatic decisions that will have applicability across the DOE complex and describes in detail the purpose and need for this DOE action. Volume 2 is specific to actions at the INEL. This document, which limits its discussion to the Savannah River Site (SRS) spent nuclear fuel management program, supports Volume 1 of the EIS. Following the introduction, Chapter 2 contains background information related to the SRS and the framework of environmental regulations pertinent to spent nuclear fuel management. Chapter 3 identifies spent nuclear fuel management alternatives that DOE could implement at the SRS, and summarizes their potential environmental consequences. Chapter 4 describes the existing environmental resources of the SRS that spent nuclear fuel activities could affect. Chapter 5 analyzes in detail the environmental consequences of each spent nuclear fuel management alternative and describes cumulative impacts. The chapter also contains information on unavoidable adverse impacts, commitment of resources, short-term use of the environment and mitigation measures.

  9. Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site

  10. Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site.

  11. Transportation of failed or damaged foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Messick, C.E. [U.S. Department of Energy, Savannah River Site (United States); Mustin, T.P. [U.S. Department of Energy (United States); Massey, C.D. [Sandia National Laboratorier (United States)

    1998-07-01

    Since resuming the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy (DOE) and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues associated with the transport of Materials Testing Reactor (MTR)-type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be important to foreign research reactor operators, shippers, and cask vendors, so that appropriate amendments to the Certificate of Compliance for spent fuel casks can be submitted in a timely manner to facilitate the safe and scheduled transport of FRR SNF.

  12. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs, Draft Environmental Impact Statement. Volume 1, Appendix D: Part A, Naval Spent Nuclear Fuel Management

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    Volume 1 to the Department of Energy`s Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Management Programs Environmental Impact Statement evaluates a range of alternatives for managing naval spent nuclear fuel expected to be removed from US Navy nuclear-powered vessels and prototype reactors through the year 2035. The Environmental Impact Statement (EIS) considers a range of alternatives for examining and storing naval spent nuclear fuel, including alternatives that terminate examination and involve storage close to the refueling or defueling site. The EIS covers the potential environmental impacts of each alternative, as well as cost impacts and impacts to the Naval Nuclear Propulsion Program mission. This Appendix covers aspects of the alternatives that involve managing naval spent nuclear fuel at four naval shipyards and the Naval Nuclear Propulsion Program Kesselring Site in West Milton, New York. This Appendix also covers the impacts of alternatives that involve examining naval spent nuclear fuel at the Expended Core Facility in Idaho and the potential impacts of constructing and operating an inspection facility at any of the Department of Energy (DOE) facilities considered in the EIS. This Appendix also considers the impacts of the alternative involving limited spent nuclear fuel examinations at Puget Sound Naval Shipyard. This Appendix does not address the impacts associated with storing naval spent nuclear fuel after it has been inspected and transferred to DOE facilities. These impacts are addressed in separate appendices for each DOE site.

  13. A study for collaborative management for nuclear spent fuel control. Seeking for nuclear non-proliferation in East Asia

    International Nuclear Information System (INIS)

    Because of the rapid increase of power generation with nuclear fuel in East Asia area, the management and control of nuclear spent fuel from nuclear reactors has become an essential and urgent issue in this area. This study focused on the possibility of forming an intergovernmental collaborative management system for nuclear spent fuel with an emphasize on nuclear non-proliferation among East Asian countries, i.e. China, Korea, Taiwan and Japan who own and operate nuclear power plants. First, we studied the present situation for nuclear spent fuel, including the storage measures, the future fore- cast on the accumulation and the government measures to deal with these spent fuel. Then, based upon first step studies, we examined the pros and cons when the collaborative management is realized particularly from the viewpoint of prevention of nuclear proliferation. Further, we estimated possible means for management and control of nuclear spent fuel, including its system size and cost. Finally, we extracted some technological tasks to be solved and political issues to be discussed. Our findings are as follows. 1. The total amount of the power generation in three East Asian counties (China, Korea and Taiwan) is about 17 million KW presently. This will be tripled to 51 million KW by the year 2010. When Japan's ability is added it is 62 million KW currently and 121 million by 2010. 2. The nuclear spent fuel in Taiwan and Korea will be saturated for their storage capacity. On the other hand, Japan will start to operate her reprocessing plant in Aomori prefecture in 2003 and her new storage capability is completed in 1999. Also in China, a reprocessing pilot plant is under construction and its operation is scheduled in 2001. 3. As their national policy, China and Japan does reprocess from spent fuel but Korea and Taiwan don't. Instead, they take non-reprocessing and direct geological disposal. 4. If the collaborative management of nuclear wastes is realized Multi

  14. Chemical toxicity in final disposal of spent nuclear fuel; Kemiallinen myrkyllisyys kaeytetyn ydinpolttoaineen loppusijoituksessa

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, E. [Posiva Oy, Helsinki (Finland); Nordman, H. [VTT Energy, Espoo (Finland)

    1999-03-01

    This report studies the chemical long-term environmental effects of the substances released from the canisters of spent nuclear fuel. An overall inventory has been made of all the chemical elements contained in the canister materials and spent fuel. The release and transport of the most toxic and abundant elements have been studied with calculations using the solubilities of elements or with analyses carried out by computer codes. The resulting concentrations of elements in groundwater have been compared to the concentration limits defined in the pertinent authority requirements for drinking water. The results show that the chemical toxicity is not a significant factor as regards the safety of the final disposal of spent nuclear fuel. (orig.) 26 refs. Working report

  15. Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

    CERN Document Server

    Poulson, D; Guardincerri, E; Morris, C L; Bacon, J D; Plaud-Ramos, K; Morley, D; Hecht, A

    2016-01-01

    Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, the cask contents can be confirmed with high confidence in less than two days exposure. Similar results can be obtained by moving a smaller detector to view the cask from multiple angles.

  16. Spent Nuclear Fuel (SNF) Project Acceptance Criteria for Light Water Reactor Spent Fuel Storage System [OCRWM PER REV2

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSON, D.M.

    2000-12-20

    As part of the decommissioning of the 324 Building Radiochemical Engineering Cells there is a need to remove commercial Light Water Reactor (LWR) spent nuclear fuel (SNF) presently stored in these hot cells. To enable fuel removal from the hot cells, the commercial LWR SNF will be packaged and shipped to the 200 Area Interim Storage Area (ISA) in a manner that satisfies site requirements for SNF interim storage. This document identifies the criteria that the 324 Building Radiochemical Engineering Cell Clean-out Project must satisfy for acceptance of the LWR SNF by the SNF Project at the 200 Area ISA. In addition to the acceptance criteria identified herein, acceptance is contingent on adherence to applicable Project Hanford Management Contract requirements and procedures in place at the time of work execution.

  17. Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

    Science.gov (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment.

  18. Radiation Effects on Materials Used in Geological Repositories for Spent Nuclear Fuel

    OpenAIRE

    Mats Jonsson

    2012-01-01

    Safe long-term storage of radioactive waste from nuclear power plants is one of the main concerns for the nuclear industry as well as for governments in countries relying on electricity produced by nuclear power. A repository for spent nuclear fuel must be safe for extremely long time periods (at least 100 000 years). In order to ascertain the long-term safety of a repository, extensive safety analysis must be performed. One of the critical issues in a safety analysis is the long-term integri...

  19. Resonance self-indication as a method for estimating fissile elements in spent nuclear assemblies

    International Nuclear Information System (INIS)

    The integral version of the neutron resonance transmission analysis (NRTA) has been developed in order to solve the problem of estimating fissile elements in spent nuclear fuel. Two variants of the integral neutron resonance transmission analysis (the resonance self-indication (the average transmission by fission) and the average full transmission) have been described

  20. Status of DOE efforts to renew acceptance of foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Head, C.R.

    1997-08-01

    This presentation summarizes the efforts being made by the Department of Energy to renew acceptance of spent nuclear fuel shipments from foreign research reactors. The author reviews the actions undertaken in this process in a fairly chronological manner, through the present time, as well as the development of an environmental impact statement to support the proposed actions.

  1. Fission product partitioning in aerosol release from simulated spent nuclear fuel

    NARCIS (Netherlands)

    Di Lemma, F.G.; Colle, J.Y.; Rasmussen, G.; Konings, R.J.M.

    2015-01-01

    Aerosols created by the vaporization of simulated spent nuclear fuel (simfuel) were produced by laser heating techniques and characterised by a wide range of post-analyses. In particular attention has been focused on determining the fission product behaviour in the aerosols, in order to improve the

  2. Spent-fuel-storage studies at the Barnwell Nuclear Fuel Plant. Studies and research concerning BNFP

    International Nuclear Information System (INIS)

    This report contains the results of various studies and demonstrations related to advanced spent-fuel-storage techniques which were performed at the Barnwell Nuclear Fuel Plant (BNFP) in 1982. The demonstrations evaluated various technical aspects of fuel disassembly and canning and dry-storage techniques. The supporting studies examined thermal limitations and criticality concerns

  3. Neutron spectrometry at the interim storage facility for spent nuclear fuel

    CERN Document Server

    Králik, M; Studeny, J

    2002-01-01

    Dosimetric characteristics of neutron and photon components of mixed fields around casks for spent nuclear fuel have been determined at various places at the dry interim storage facility. The results obtained with metrological grade instruments were compared with data provided by usual survey meters for both neutrons and photons.

  4. Spent Nuclear Fuel Management: Levelized Cost of Electricity Generation and Analysis of Various Production Scenarios

    Directory of Open Access Journals (Sweden)

    Laura Rodriguez-Penalonga

    2016-03-01

    Full Text Available This article aims to analyze the results of an economic study carried out to compare the influence of nuclear production capacity in different countries. The analysis is based on LCOEs (levelized cost of electricity for three back-end strategies: open cycle, closed cycle and advanced closed cycle. The results show that costs are not a relevant criteria in order to select an energy policy for the spent nuclear fuel management.

  5. German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

    OpenAIRE

    SCHWENK-FERRERO A.

    2013-01-01

    Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the...

  6. Literature on fabrication of tungsten for application in pyrochemical processing of spent nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Edstrom, C.M.; Phillips, A.G.; Johnson, L.D.; Corle, R.R.

    1980-10-11

    The pyrochemical processing of nuclear fuels requires crucibles, stirrers, and transfer tubing that will withstand the temperature and the chemical attack from molten salts and metals used in the process. This report summarizes the literature that pertains to fabrication (joining, chemical vapor deposition, plasma spraying, forming, and spinning) is the main theme. This report also summarizes a sampling of literature on molbdenum and the work previously performed at Argonne National Laboratory on other container materials used for pyrochemical processing of spent nuclear fuels.

  7. Literature on fabrication of tungsten for application in pyrochemical processing of spent nuclear fuels

    International Nuclear Information System (INIS)

    The pyrochemical processing of nuclear fuels requires crucibles, stirrers, and transfer tubing that will withstand the temperature and the chemical attack from molten salts and metals used in the process. This report summarizes the literature that pertains to fabrication (joining, chemical vapor deposition, plasma spraying, forming, and spinning) is the main theme. This report also summarizes a sampling of literature on molbdenum and the work previously performed at Argonne National Laboratory on other container materials used for pyrochemical processing of spent nuclear fuels

  8. SFCOMPO: A new database of isotopic compositions of spent nuclear fuel

    International Nuclear Information System (INIS)

    The numerous applications of nuclear fuel depletion simulations impact all areas related to nuclear safety. They are at the basis of, inter alia, spent fuel criticality safety analyses, reactor physics calculations, burn-up credit methodologies, decay heat thermal analyses, radiation shielding, reprocessing, waste management, deep geological repository safety studies and safeguards. Experimentally determined nuclide compositions of well-characterised spent nuclear fuel (SNF) samples are used to validate the accuracy of depletion code predictions for a given burn-up. At the same time, the measured nuclide composition of the sample is used to determine the burn-up of the fuel. It is therefore essential to have a reliable and well-qualified database of measured nuclide concentrations and relevant reactor operational data that can be used as experimental benchmark data for depletion codes and associated nuclear data. The Spent Fuel Isotopic Composition Database (SFCOMPO) has been hosted by the NEA since 2001. In 2012, a collaborative effort led by the NEA Data Bank and Oak Ridge National Laboratory (ORNL) in the United States, under the guidance of the NEA Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF) of the Working Party on Nuclear Criticality Safety (WPNCS), has resulted in the creation of an enhanced relational database structure and a significant expansion of the SFCOMPO database, which now contains experimental assay data for a wider selection of international reactor designs. The new database was released online in 2014. This new SFCOMPO database aims to provide access to open experimental SNF assay data to ensure their preservation and to facilitate their qualification as evaluated assay data suitable for the validation of methodologies used to predict the composition of irradiated nuclear fuel. Having a centralised, internationally reviewed database that makes these data openly available for a large selection of international reactor designs is of

  9. Safeguards for final disposal of spent nuclear fuel. Methods and technologies for the Olkiluoto site

    International Nuclear Information System (INIS)

    The final disposal of the nuclear material shall introduce new safeguards concerns which have not been addressed previously in IAEA safeguards approaches for spent fuel. The encapsulation plant to be built at the site will be the final opportunity for verification of spent fuel assemblies prior to their transfer to the geological repository. Moreover, additional safety and safeguards measures are considered for the underground repository. Integrated safeguards verification systems will also concentrate on environmental monitoring to observe unannounced activities related to possible diversion schemes at the repository site. The final disposal of spent nuclear fuel in geological formation will begin in Finland within 10 years. After the geological site investigations and according to legal decision made in 2001, the final repository of the spent nuclear fuel shall be located at the Olkiluoto site in Eurajoki. The next phase of site investigations contains the construction of an underground facility, called ONKALO, for rock characterisation purposes. The excavation of the ONKALO is scheduled to start in 2004. Later on, the ONKALO may form a part of the final repository. The plans to construct the underground facility for nuclear material signify that the first safeguards measures, e.g. baseline mapping of the site area, need to take prior to the excavation phase. In order to support the development and implementation of the regulatory control of the final disposal programme, STUK established an independent expert group, LOSKA. The group should support the STUK in the development of the technical safeguards requirements, in the implementation of the safeguards and in the evaluation of the plans of the facility operator. This publication includes four background reports produced by this group. The first of these 'NDA verification of spent fuel, monitoring of disposal canisters, interaction of the safeguards and safety issues in the final disposal' describes the new

  10. Criticality safety aspects of spent fuel arrays from emerging nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Nicolaou, G. [University of Thrace, Department of Electrical and Computer Engineering, Laboratory of Nuclear Technology, Kimmerria Campus, 67100 Xanthi (Greece)

    2010-07-01

    Emerging nuclear fuel cycles: fuels with Pu or minor actinides (MA) for their self-generated recycling or transmutation in PWR or FR {yields} reduction of radiotoxicity of HLW. The aim of work is to assess criticality (k{sub {infinity}}) of arrays of spent nuclear fuels from these emerging fuel cycles. Procedures: Calculations of - k{sub {infinity}}, using MCNP5 based on fresh and spent fuel compositions (infinite arrays), - spent fuel compositions using ORIGEN. Fuels considered: - commercial PWR-UO{sub 2} (R1) and -MOX (R2), [45 GWd/t] and fast reactor [100 GWd/t] (R3), - PWR self-generated Pu recycling (S1) and MA recycling (S2), FR self-generated MA recycling (S3), FR with 2% {sup 237}Np for transmutation purposes (T). Results: k{sub {infinity}} based on fresh and spent fuel compositions is shown. Fuels are clustered in two distinct families: - fast reactor fuels, - thermal reactor fuels; k{sub {infinity}} decreases when calculated on the basis of actinide and fission product inventory. In conclusions: - Emerging fuels considered resemble their corresponding commercial fuels; - k{sub {infinity}} decreases in all cases when calculated on the basis of spent fuel compositions (reactivity worth {approx}-20%{Delta}k/k), hence improving the effectiveness of packaging. (author)

  11. Storage facilities of spent nuclear fuel in dry for Mexican nuclear facilities; Instalaciones de almacenamiento de combustible nuclear gastado en seco para instalaciones nucleares mexicanas

    Energy Technology Data Exchange (ETDEWEB)

    Salmeron V, J. A.; Camargo C, R.; Nunez C, A.; Mendoza F, J. E.; Sanchez J, J., E-mail: juan.salmeron@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    In this article the relevant aspects of the spent fuel storage and the questions that should be taken in consideration for the possible future facilities of this type in the country are approached. A brief description is proposed about the characteristics of the storage systems in dry, the incorporate regulations to the present Nuclear Regulator Standard, the planning process of an installation, besides the approaches considered once resolved the use of these systems; as the modifications to the system, the authorization periods for the storage, the type of materials to store and the consequent environmental impact to their installation. At the present time the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) considers the possible generation of two authorization types for these facilities: Specific, directed to establish a new nuclear installation with the authorization of receiving, to transfer and to possess spent fuel and other materials for their storage; and General, focused to those holders that have an operation license of a reactor that allows them the storage of the nuclear fuel and other materials that they possess. Both authorizations should be valued according to the necessities that are presented. In general, this installation type represents a viable solution for the administration of the spent fuel and other materials that require of a temporary solution previous to its final disposal. Its use in the nuclear industry has been increased in the last years demonstrating to be appropriate and feasible without having a significant impact to the health, public safety and the environment. Mexico has two main nuclear facilities, the nuclear power plant of Laguna Verde of the Comision Federal de Electricidad (CFE) and the facilities of the TRIGA Reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) that will require in a future to use this type of disposition installation of the spent fuel and generated wastes. (Author)

  12. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs draft environmental impact statement. Volume 1, Appendix B: Idaho National Engineering Laboratory Spent Nuclear Fuel Management Program

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The US Department of Energy (DOE) has prepared this report to assist its management in making two decisions. The first decision, which is programmatic, is to determine the management program for DOE spent nuclear fuel. The second decision is on the future direction of environmental restoration, waste management, and spent nuclear fuel management activities at the Idaho National Engineering Laboratory. Volume 1 of the EIS, which supports the programmatic decision, considers the effects of spent nuclear fuel management on the quality of the human and natural environment for planning years 1995 through 2035. DOE has derived the information and analysis results in Volume 1 from several site-specific appendixes. Volume 2 of the EIS, which supports the INEL-specific decision, describes environmental impacts for various environmental restoration, waste management, and spent nuclear fuel management alternatives for planning years 1995 through 2005. This Appendix B to Volume 1 considers the impacts on the INEL environment of the implementation of various DOE-wide spent nuclear fuel management alternatives. The Naval Nuclear Propulsion Program, which is a joint Navy/DOE program, is responsible for spent naval nuclear fuel examination at the INEL. For this appendix, naval fuel that has been examined at the Naval Reactors Facility and turned over to DOE for storage is termed naval-type fuel. This appendix evaluates the management of DOE spent nuclear fuel including naval-type fuel.

  13. Separation of the rare-earth fission product poisons from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Christian, Jerry D.; Sterbentz, James W.

    2016-08-30

    A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO.sub.2 and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO.sub.2 in the nuclear fuel, preferably to at least to 850.degree. C. for Th and up to 600.degree. C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.

  14. Temporary storage in dry of the spent nuclear fuel in the Nuclear Power Plant of Laguna Verde

    International Nuclear Information System (INIS)

    To guarantee the continuity in the operation of the two nuclear reactors of the nuclear power plant of Laguna Verde (NPP-L V) is an activity of high priority of the Comision Federal de Electricidad (CFE) in Mexico. At the present time, the CFE is working in the storage project in dry of the spent fuel with the purpose of to liberate space of the pools and to have the enlarged capacity of storage of the spent fuel that is discharged of the reactors. This work presents the storage option in dry of the spent fuel, considering that the original capacity of the spent fuel pools of the NPP-L V was of 1242 spaces each one and that in 1991, through a modification of the original design, the storage capacity was increased to 3177 spaces by pool. At present, the cells occupied by unit are of 2165 (68%) for the Unit-I and 1839 (58%) for the Unit-2, however, in 2017 and 2022 the capacity to discharge the complete core will be limited by what is required of a retirement option of spent fuel assemblies to liberate spaces. (author)

  15. Research and development for decontamination system of spent resin in Hanbit Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Gi Hong [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2015-12-15

    When reactor coolant leaks occur due to cracks of a steam generator tube, radioactive materials contained in the primary cooling water in nuclear power plant are forced out toward the secondary systems. At this time the secondary water purification resin in the ion exchange resin tower of the steam generator blowdown system is contaminated by the radioactivity of the leaked radioactive materials, so we pack this in special containers and store temporarily because we could not dispose it by ourselves. If steam generator tube leakage occurs, it produces contaminated spent resins annually about 5,000-7,000 liters. This may increase the amount of nuclear waste productions, a disposal working cost and a unit price of generating electricity in the plant. For this reasons, it is required to develop a decontamination process technique for reducing the radioactive level of these resins enough to handle by the self-disposal method. In this research, First, Investigated the structure and properties of the ion exchange resin used in a steam generator blowdown system. Second, Checked for a occurrence status of contaminated spent resin and a disposal technology. Third, identified the chemical characteristics of the waste radionuclides of the spent resin, and examined ionic bonding and separation mechanism of radioactive nuclear species and a spent resin. Finally, we carried out the decontamination experiment using chemicals, ultrasound, microbubbles, supercritical carbon dioxide to process these spent resin. In the case of the spent resin decontamination method using chemicals, the higher the concentration of the drug decontamination efficiency was higher. In the ultrasound method, foreign matter of the spent resin was removed and was found that the level of radioactivity is below of the MDA. In the microbubbles method, we found that the concentration of the radioactivity decreased after the experiment, so it can be used to the decontamination process of the spent resin. In

  16. Research on advanced aqueous reprocessing of spent nuclear fuel: literature study

    International Nuclear Information System (INIS)

    The goal of the partitioning and transmutation strategy is to reduce the radiotoxicity of spent nuclear fuel to the level of natural uranium in a short period of time (about 1000 years) and thus the required containment period of radioactive material in a repository. Furthermore, it aims to reduce the volume of waste requiring deep geological disposal and hence the associated space requirements and costs. Several aqueous as well as pyrochemical separation processes have been developed for the partitioning of the long-lived radionuclides from the remaining of the spent fuel. This report aims to describe and compare advanced aqueous reprocessing methods.

  17. Research on advanced aqueous reprocessing of spent nuclear fuel: literature study

    Energy Technology Data Exchange (ETDEWEB)

    Van Hecke, K.; Goethals, P.

    2006-07-15

    The goal of the partitioning and transmutation strategy is to reduce the radiotoxicity of spent nuclear fuel to the level of natural uranium in a short period of time (about 1000 years) and thus the required containment period of radioactive material in a repository. Furthermore, it aims to reduce the volume of waste requiring deep geological disposal and hence the associated space requirements and costs. Several aqueous as well as pyrochemical separation processes have been developed for the partitioning of the long-lived radionuclides from the remaining of the spent fuel. This report aims to describe and compare advanced aqueous reprocessing methods.

  18. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  19. A study on Japanese experience to secure the interim storage facility for nuclear spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Min [Hannyang University, Seoul (Korea, Republic of)

    2007-12-15

    The Japanese Government selected Mutsu, Aomori Prefecture as a provisional spent-fuel repository site. This comes as a result of the prefecture's five-year campaign to host the site since 2000. Korea stores spent nuclear fuel within sites of nuclear power plants, and expects the storage capacity to reach its limit by the year 2016. This compels Korea to learn the cases of Japan. Having successfully hosted Gyeongju as a site for low-to-intermediate-level nuclear waste repository, Korea has already learned the potential process of hosting spent fuel storage site. The striking difference between the two countries in the process of hosting the site is that the Korean government had to offer the local city a large amount of subsidy for hosting through competitive citizens' referendum among candidate cities while it was the leadership of the local municipality that enabled the controversial decision in Japan. It is also a distinguishable characteristics of Japan that not a huge subsidy is provided to the local host city. I hope this study offers an idea to Korea's future effort to select a spent-fuel host site.

  20. A new OECD/NEA database of nuclide compositions of spent nuclear fuel

    International Nuclear Information System (INIS)

    The SFCOMPO database of nuclide compositions of spent nuclear fuel is hosted by the OECD Nuclear Energy Agency since 2001. Since 2011, a collaborative effort led by the OECD/NEA Data Bank and Oak Ridge National Laboratory, under the guidance of the NEA Expert Group on Assay Data of Spent Nuclear Fuel, has resulted in the creation of a new enhanced relational database structure and a significant expansion of SFCOMPO, now containing experimental assay data for a wide selection of international reactor designs. This paper aims at describing the new SFCOMPO Database developed at NEA in terms of functionalities, contents and foreseen developments. This new database is expected for public release in 2014. (author)

  1. Modelling the inventory and impact assessment of partitioning and transmutation approaches to spent nuclear fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Hoggett-Jones, C. E-mail: craig@stams.strath.ac.uk; Robbins, C.; Gettinby, G.; Blythe, S

    2002-03-01

    An inventory modelling and impact assessment system to investigate the potential effects of partitioning and transmutation is proposed. It is founded on a mass based inventory analysis using the principles of basic nuclear physics and the international standards for assessing radiological health effects. It is specific to the back-end of the nuclear fuel cycle and is applied to four alternative spent fuel management strategies. The system accounts for the dynamic nature of post-irradiation scenarios and is being used to develop software for use within the nuclear power industry. Four example waste-disposal options are considered using the method. Impact assessments and parameter sensitivity analyses are presented.

  2. Modelling the inventory and impact assessment of partitioning and transmutation approaches to spent nuclear fuel management

    International Nuclear Information System (INIS)

    An inventory modelling and impact assessment system to investigate the potential effects of partitioning and transmutation is proposed. It is founded on a mass based inventory analysis using the principles of basic nuclear physics and the international standards for assessing radiological health effects. It is specific to the back-end of the nuclear fuel cycle and is applied to four alternative spent fuel management strategies. The system accounts for the dynamic nature of post-irradiation scenarios and is being used to develop software for use within the nuclear power industry. Four example waste-disposal options are considered using the method. Impact assessments and parameter sensitivity analyses are presented

  3. Application of system analysis to the solution of economic problems of spent nuclear fuels in Poland

    International Nuclear Information System (INIS)

    In terms of unit capacity of spent fuel reprocessing plants and timing of their introduction, the problem of optimizing the fuel reprocessing industry has been investigated for the given growth rates of nuclear electricity generating capacities and with a special reference to Poland. The summary of two approaches to solve the problem is presented. The first one utilizes the Bellman dynamic programming method and determines the optimal timing scheme of putting reprocessing plants into operation with an account of a possibility for a part of the spent fuel to be reprocessed abroad. The second approach utilized the static optimization method. It is pointed out that partial reprocessing of spent fuel abroad may save a significant part of the total reprocessing costs especially for small economics like Poland

  4. Performance assessment of self-interrogation neutron resonance densitometry for spent nuclear fuel assay

    Science.gov (United States)

    Hu, Jianwei; Tobin, Stephen J.; LaFleur, Adrienne M.; Menlove, Howard O.; Swinhoe, Martyn T.

    2013-11-01

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is one of several nondestructive assay (NDA) techniques being integrated into systems to measure spent fuel as part of the Next Generation Safeguards Initiative (NGSI) Spent Fuel Project. The NGSI Spent Fuel Project is sponsored by the US Department of Energy's National Nuclear Security Administration to measure plutonium in, and detect diversion of fuel pins from, spent nuclear fuel assemblies. SINRD shows promising capability in determining the 239Pu and 235U content in spent fuel. SINRD is a relatively low-cost and lightweight instrument, and it is easy to implement in the field. The technique makes use of the passive neutron source existing in a spent fuel assembly, and it uses ratios between the count rates collected in fission chambers that are covered with different absorbing materials. These ratios are correlated to key attributes of the spent fuel assembly, such as the total mass of 239Pu and 235U. Using count rate ratios instead of absolute count rates makes SINRD less vulnerable to systematic uncertainties. Building upon the previous research, this work focuses on the underlying physics of the SINRD technique: quantifying the individual impacts on the count rate ratios of a few important nuclides using the perturbation method; examining new correlations between count rate ratio and mass quantities based on the results of the perturbation study; quantifying the impacts on the energy windows of the filtering materials that cover the fission chambers by tallying the neutron spectra before and after the neutrons go through the filters; and identifying the most important nuclides that cause cooling-time variations in the count rate ratios. The results of these studies show that 235U content has a major impact on the SINRD signal in addition to the 239Pu content. Plutonium-241 and 241Am are the two main nuclides responsible for the variation in the count rate ratio with cooling time. In short, this work

  5. Investigation of Electrochemical Recovery of Zirconium from Spent Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Michael Simpson; II-Soon Hwang

    2014-06-01

    This project uses both modeling and experimental studies to design optimal electrochemical technology methods for recovery of zirconium from used nuclear fuel rods for more effective waste management. The objectives are to provide a means of efficiently separating zirconium into metallic high-level waste forms and to support development of a process for decontamination of zircaloy hulls to enable their disposal as low- and intermediate-level waste. Modeling work includes extension of a 3D model previously developed by Seoul National University for uranium electrorefining by adding the ability to predict zirconium behavior. Experimental validation activities include tests for recovery of zirconium from molten salt solutions and aqueous tests using surrogate materials. *This is a summary of the FY 2013 progress for I-NERI project # 2010-001-K provided to the I-NERI office.

  6. Spent Fuel Challenges Facing Small and New Nuclear Programmes

    International Nuclear Information System (INIS)

    In order to ensure that the radioactive wastes in any country are managed safely, it is necessary to have an established legislative and regulatory framework and also to create the necessary organizations for implementation and for oversight of waste management operations and facility development. Guidance on these issues is given in the Joint Convention and a number of other IAEA documents. The IAEA, and also the EC, have in addition published key overarching strategic advisory documents for new nuclear programmes. These tend to imply that all nuclear programmes, however large or small, should be pressing ahead urgently towards early implementation of geological repositories. In practice, however, in small programmes there are neither economic nor technical drivers for early implementation of deep geological repositories; constructing simpler facilities for the disposal of the larger volume of low-level waste has higher priority. Nevertheless, in all countries political decisions have to be taken and policies set in place to ensure that geological disposal will implemented without unjustified delay. This paper distils out a set of key messages for small programmes. Amongst the most critical are the following. Even if disposal is far off, planning and organization should begin at the initiation of the programme; this can help with technical and economic optimization and (importantly) also with public and political acceptance. Important lessons can be learned from advanced programmes — but these must be adapted to allow for the different boundary conditions of new and small programmes. The key differences relate to the timescales involved, and the resources available. There is a range of waste management and waste disposal options open to new programmes. It is not necessary to choose definitive solutions at the outset; options can be kept open, but a minimum level of engagement is required for all open options. (author)

  7. REACTOR PHYSICS MODELING OF SPENT NUCLEAR RESEARCH REACTOR FUEL FOR SNM ATTRIBUTION AND NUCLEAR FORENSICS

    Energy Technology Data Exchange (ETDEWEB)

    Sternat, M.; Beals, D.; Webb, R.; Nichols, T.

    2010-06-09

    Nuclear research reactors are the least safeguarded type of reactor; in some cases this may be attributed to low risk and in most cases it is due to difficulty from dynamic operation. Research reactors vary greatly in size, fuel type, enrichment, power and burnup providing a significant challenge to any standardized safeguard system. If a whole fuel assembly was interdicted, based on geometry and other traditional forensics work, one could identify the material's origin fairly accurately. If the material has been dispersed or reprocessed, in-depth reactor physics models may be used to help with the identification. Should there be a need to attribute research reactor fuel material, the Savannah River National Laboratory would perform radiochemical analysis of samples of the material as well as other non-destructive measurements. In depth reactor physics modeling would then be performed to compare to these measured results in an attempt to associate the measured results with various reactor parameters. Several reactor physics codes are being used and considered for this purpose, including: MONTEBURNS/ORIGEN/MCNP5, CINDER/MCNPX and WIMS. In attempt to identify reactor characteristics, such as time since shutdown, burnup, or power, various isotopes are used. Complexities arise when the inherent assumptions embedded in different reactor physics codes handle the isotopes differently and may quantify them to different levels of accuracy. A technical approach to modeling spent research reactor fuel begins at the assembly level upon acquiring detailed information of the reactor to be modeled. A single assembly is run using periodic boundary conditions to simulate an infinite lattice which may be repeatedly burned to produce input fuel isotopic vectors of various burnups for a core level model. A core level model will then be constructed using the assembly level results as inputs for the specific fuel shuffling pattern in an attempt to establish an equilibrium cycle

  8. Assessment of the risk of transporting spent nuclear fuel by truck

    Energy Technology Data Exchange (ETDEWEB)

    Elder, H.K.

    1978-11-01

    The assessment includes the risks from release of spent fuel materials and radioactive cask cavity cooling water due to transportation accidents. The contribution to the risk of package misclosure and degradation during normal transport was also considered. The results of the risk assessment have been related to a time in the mid-1980's, when it is projected that nuclear plants with an electrical generating capacity of 100 GW will be operating in the U.S. For shipments from reactors to interim storage facilities, it is estimated that a truck carrying spent fuel will be involved in an accident that would not be severe enough to result in a release of spent fuel material about once in 1.1 years. It was estimated that an accident that could result in a small release of radioactive material (primarily contaminated cooling water) would occur once in about 40 years. The frequency of an accident resulting in one or more latent cancer fatalities from release of radioactive materials during a truck shipment of spent fuel to interim storage was estimated to be once in 41,000 years. No accidents were found that would result in acute fatalities from releases of radioactive material. The risk for spent fuel shipments from reactors to reprocessing plants was found to be about 20% less than the risk for shipments to interim storage. Although the average shipment distance for the reprocessing case is larger, the risk is somewhat lower because the shipping routes, on average, are through less populated sections of the country. The total risk from transporting 180-day cooled spent fuel by truck in the reference year is 4.5 x 10/sup -5/ fatalities. An individual in the population at risk would have one chance in 6 x 10/sup 11/ of suffering a latent cancer fatality from a release of radioactive material from a truck carrying spent fuel in the reference year. (DLC)

  9. Assessment of the risk of transporting spent nuclear fuel by truck

    International Nuclear Information System (INIS)

    The assessment includes the risks from release of spent fuel materials and radioactive cask cavity cooling water due to transportation accidents. The contribution to the risk of package misclosure and degradation during normal transport was also considered. The results of the risk assessment have been related to a time in the mid-1980's, when it is projected that nuclear plants with an electrical generating capacity of 100 GW will be operating in the U.S. For shipments from reactors to interim storage facilities, it is estimated that a truck carrying spent fuel will be involved in an accident that would not be severe enough to result in a release of spent fuel material about once in 1.1 years. It was estimated that an accident that could result in a small release of radioactive material (primarily contaminated cooling water) would occur once in about 40 years. The frequency of an accident resulting in one or more latent cancer fatalities from release of radioactive materials during a truck shipment of spent fuel to interim storage was estimated to be once in 41,000 years. No accidents were found that would result in acute fatalities from releases of radioactive material. The risk for spent fuel shipments from reactors to reprocessing plants was found to be about 20% less than the risk for shipments to interim storage. Although the average shipment distance for the reprocessing case is larger, the risk is somewhat lower because the shipping routes, on average, are through less populated sections of the country. The total risk from transporting 180-day cooled spent fuel by truck in the reference year is 4.5 x 10-5 fatalities. An individual in the population at risk would have one chance in 6 x 1011 of suffering a latent cancer fatality from a release of radioactive material from a truck carrying spent fuel in the reference year

  10. Implementation of burnup and control rod credit for storage of spent nuclear fuel in Ukraine

    International Nuclear Information System (INIS)

    Preliminary analysis of the regulations in force in Ukraine concerning nuclear safety of spent nuclear fuel management systems shows that some regulatory requirements in force are too conservative in view of current international practice. The extent of conservatism can be determined and reduced, if necessary, only using calculated studies for analyzing the criticality of spent nuclear fuel management systems. Such activity is consistent with the requirements posed by state-of-the-art production requirements. However, this can be only based on improving our level of understanding the processes occurring in nuclear dangerous systems and improving our capabilities as regards accuracy, correctness, and reliability in numerical modeling these processes. This work was intended to demonstrate that the excessive conservatism laid previously into the requirements on nuclear safety in Ukraine due to insufficient development of means for modeling processes in nuclear fuel can be considerably decreased through using more real modeling fuel systems. If such modeling is performed with the use of state-of-the-art software and computers, based on more complete understanding the processes in fuel systems, then removal of the excessive conservatism does will not reduce the safety of nuclear dangerous systems. (author)

  11. The State of the Art of the Decladding Method for the Spent Nuclear Fuel Rods

    International Nuclear Information System (INIS)

    Our country's energy consumption is increasing day after day even though it relies on imports for more than 95 percent of its energy needs. In this circumstances, the atomic energy is a promising alternative to solve the problem of an energy security and an environmental preservation simultaneously. However, nuclear power produces spent fuel which is a highly radioactive waste. For a reliable and effective management of the spent fuel, the ACP(Advanced Spent Fuel Conditioning Process) is being developed at the KAERI. As a first state equipment of the ACP, a decladding machine is used to separate spent fuel rod into the UO2 pellets and hulls. This technical report aims to analyze existing decladding methods, and then, find a suitable decladding mechanism for the ACP. Many studies on the decladding of spent fuel can be categorized two approaches: chemical approach and mechanical one. In this report, we concentrated on the mechanical decladding approach. We developed engineering scale decladding device(20 kgHM/batch) and evaluated the performance through the verification experiments. We expect that this technical report helps in developing a scale-up equipment and technology

  12. Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions

  13. Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

  14. Safety case for the disposal of spent nuclear fuel at Olkiluoto - Synthesis 2012

    International Nuclear Information System (INIS)

    TURVA-2012 is Posiva's safety case in support of the Preliminary Safety Analysis Report (PSAR 2012) and application for a construction licence for a spent nuclear fuel repository. Consistent with the Government Decisions-in- Principle, this foresees a repository developed in bedrock at the Olkiluoto site according to the KBS-3 method, designed to accept spent nuclear fuel from the lifetime operations of the Olkiluoto and Loviisa reactors. Synthesis 2012 presents a synthesis of Posiva Oy's Safety Case 'TURVA-2012' portfolio. It summarises the design basis for the repository at the Olkiluoto site, the assessment methodology and key results of performance and safety assessments. It brings together all the lines of argument for safety, evaluation of compliance with the regulatory requirements, and statement of confidence in long-term safety and Posiva's safety analyses. The TURVA-2012 safety case demonstrates that the proposed repository design provides a safe solution for the disposal of spent nuclear fuel, and that the performance and safety assessments are fully consistent with all the legal and regulatory requirements related to long-term safety as set out in Government Decree 736/2008 and in guidance from the nuclear regulator - the STUK. Moreover, Posiva considers that the level of confidence in the demonstration of safety is appropriate and sufficient to submit the construction licence application to the authorities. The assessment of long-term safety includes uncertainties, but these do not affect the basic conclusions on the long-term safety of the repository. (orig.)

  15. Public relations campaign for shipping spent nuclear fuel

    International Nuclear Information System (INIS)

    An example of positive outcome of proper attitude of the media and public on the occasion of shipping of nuclear fuel is described. Nothing new was invented in the way of public relations issue management. But a combination of a number of proven techniques were put together and the public relations plan that was highly successful. early planning was of great help. Public officials were well informd by means of ANS organized seminars. ANS had experts from Sandia Labs (a major government research facility), General Electric (the cask supplier), the railroad we planned to use and Northern States Power Company on the program to describe what was going to happen and why it was safe. These sessions are believed to head off a major portion of the local opposition. A cooperation was established with the states of Wisconsin and Minnesota in providing shipment-specific training for emergency response personnel along the route. Safety, obviously, was the number one concern expressed by public officials. Knowing that would be the case, it was decided to provide some optional extras to go with the shipments. There was a consultant yo do a safety analysis of all the possible rail routes between the plant and storage facility. Though none was required by law, a shipment-specific emergency response plan which was prepared. Another important effort which was maintained from the beginning was sharing information among the participants. In dealing with the news media, an attemp was made to stick to a single source of information as much as possible. When dealing with the news media, one should refuse to apologize for modern technology. One should attack, at every opportunity, the idea that a risk-freesociety is worth the price of returning to the Dark Ages. The contributions of nuclear technology are numerous and far-reaching. Its negative impacts on health and safety have been minor compared with most other major industrial technologies. Certainly there is risk in stepping out of

  16. Spent Fuel Working Group report on inventory and storage of the Department's spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities

    International Nuclear Information System (INIS)

    A self assessment was conducted of those Hanford facilities that are utilized to store Reactor Irradiated Nuclear Material, (RINM). The objective of the assessment is to identify the Hanford inventories of RINM and the ES ampersand H concerns associated with such storage. The assessment was performed as proscribed by the Project Plan issued by the DOE Spent Fuel Working Group. The Project Plan is the plan of execution intended to complete the Secretary's request for information relevant to the inventories and vulnerabilities of DOE storage of spent nuclear fuel. The Hanford RINM inventory, the facilities involved and the nature of the fuel stored are summarized. This table succinctly reveals the variety of the Hanford facilities involved, the variety of the types of RINM involved, and the wide range of the quantities of material involved in Hanford's RINM storage circumstances. ES ampersand H concerns are defined as those circumstances that have the potential, now or in the future, to lead to a criticality event, to a worker radiation exposure event, to an environmental release event, or to public announcements of such circumstances and the sensationalized reporting of the inherent risks

  17. Seismic considerations for spent nuclear fuel storage in dry casks

    Institute of Scientific and Technical Information of China (English)

    John L Bignell; Jeffrey A Smith; Christopher A Jones; Susan Y Pickering

    2013-01-01

    To aid the United States Nuclear Regulatory Commission,Sandia National Laboratories (SNL) was contracted to investigate the seismic behavior of typical dry cask storage systems.Parametric evaluations characterized the sensitivity of calculated cask response characteristics to input parameters.The parametric evaluation investigated two generic cask designs (cylindrical and rectangular),three different foundation types (soft soil,hard soil,and rock),and three different casks to pad coefficients of friction (0.2,0.55,0.8) for earthquakes with peak ground accelerations of 0.25g,0.6g,1.0g and 1.25g.A total of 1 165 analyses were completed,with regression analyses being performed on the resulting cask response data to determine relationships relating the mean (16 % and 84 % confidence intervals on the mean) to peak ground acceleration,peak ground velocity,and pseudo-spectral acceleration at 1 Hz and 5 % damping.In general,the cylindrical casks experienced significantly larger responses in comparison to the rectangular cask.The cylindrical cask experienced larger top of cask displacements,larger cask rotations (rocking),and more occurrences of cask toppling (the rectangular cask never toppled over).The cylindrical cask was also susceptible to rolling once rocking had been initiated,a behavior not observed in the rectangular cask.Cask response was not overly sensitive to foundation type,but was significantly dependent on the response spectrum employed.

  18. Design of a Prototype Differential Die-Away Instrument Proposed for Swedish Spent Nuclear Fuel Characterization

    Science.gov (United States)

    Martinik, Tomas; Henzl, Vladimir; Grape, Sophie; Jansson, Peter; Swinhoe, Martyn T.; Goodsell, Alison V.; Tobin, Stephen J.

    2016-06-01

    As part of the United States (US) Department of Energy's Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project, the traditional Differential Die-Away (DDA) method that was originally developed for waste drum assay has been investigated and modified to provide a novel application to characterize or verify spent nuclear fuel (SNF). Following the promising, yet largely theoretical and simulation based, research of physics aspects of the DDA technique applied to SNF assay during the early stages of the NGSI-SF project, the most recent effort has been focused on the practical aspects of developing the first fully functional and deployable DDA prototype instrument for spent fuel. As a result of the collaboration among US research institutions and Sweden, the opportunity to test the newly proposed instrument's performance with commercial grade SNF at the Swedish Interim Storage Facility (Clab) emerged. Therefore the design of this instrument prototype has to accommodate the requirements of the Swedish regulator as well as specific engineering constrains given by the unique industrial environment. Within this paper, we identify key components of the DDA based instrument and we present methodology for evaluation and the results of a selection of the most relevant design parameters in order to optimize the performance for a given application, i.e. test-deployment, including assay of 50 preselected spent nuclear fuel assemblies of both pressurized (PWR) as well as boiling (BWR) water reactor type.

  19. Analysis of the risk of transporting spent nuclear fuel by train

    Energy Technology Data Exchange (ETDEWEB)

    Elder, H.K.

    1981-09-01

    This report uses risk analyses to analyze the safety of transporting spent nuclear fuel for commercial rail shipping systems. The rail systems analyzed are those expected to be used in the United States when the total electricity-generating capacity by nuclear reactors is 100 GW in the late 1980s. Risk as used in this report is the product of the probability of a release of material to the environment and the consequences resulting from the release. The analysis includes risks in terms of expected fatalities from release of radioactive materials due to transportation accidents involving PWR spent fuel shipped in rail casks. The expected total risk from such shipments is 1.3 x 10/sup -4/ fatalities per year. Risk spectrums are developed for shipments of spent fuel that are 180 days and 4 years out-of-reactor. The risk from transporting spent fuel by train is much less (by 2 to 4 orders of magnitude) than the risk to society from other man-caused events such as dam failure.

  20. Dry Storage Demonstration for High-Burnup Spent Nuclear Fuel-Feasibility Study

    International Nuclear Information System (INIS)

    Initially, casks for dry storage of spent fuel were licensed for assembly-average burnup of about 35 GWd/MTU. Over the last two decades, the discharge burnup of fuel has increased steadily and now exceeds 45 GWd/MTU. With spent fuel burnups approaching the licensing limits (peak rod burnup of 62 GWd/MTU for pressurized water reactor fuel) and some lead test assemblies being burned beyond this limit, a need for a confirmatory dry storage demonstration program was first identified after the publication in May 1999 of the U.S. Nuclear Regulatory Commissions (NRC) Interim Staff Guidance 11 (ISG-11). With the publication in July 2002 of the second revision of ISG-11, the desirability for such a program further increased to obtain confirmatory data about the potential changes in cladding mechanical properties induced by dry storage, which would have implications to the transportation, handling, and disposal of high-burnup spent fuel. While dry storage licenses have kept pace with reactor discharge burnups, transportation licenses have not and are considered on a case by case basis. Therefore, this feasibility study was performed to examine the options available for conducting a confirmatory experimental program supporting the dry storage, transportation, and disposal of spent nuclear fuel with burnups well in excess of 45 GWd/MTU

  1. Slag cement matrix for immobilisation of spent organic ion exchange resins from nuclear power plants

    International Nuclear Information System (INIS)

    The spent IX resins from nuclear power reactors are highly active solid wastes generated during operations of water cooled nuclear power plants. Direct immobilisation of spent IX resins in slag cement matrix was investigated. Slag cement matrix formulations were optimised for mixed, K+ and borate resins by evaluating leaching characteristics, porosity, hydraulic conductivity etc. The 137cesium leaching characteristics of slag cement matrices developed in this study gave the average leach rates of the order of 10-4 g/cm2/d and leachability indices of 8.4 to 9.6. The 137cesium behaviour was in accordance with the porosity and hydraulic conductivity of matrix. The laboratory study revealed that these tests can be effectively used to shortlist cement matrix formulations for long term 137cesium leach tests. The results showed that slag cement matrix can be satisfactorily used to immobilise spent resins arising from PHWR and PWR/VVER type reactors. The selected formulations were successfully deployed to prepare full size (200 litre) slag cement matrix block incorporating 100 litres of mixed IX resins. The compressive strengths of core drilled samples were 64 to 118 kg/cm2 which satisfactorily meet the safety requirements. The study has shown that the slag cement matrix has potential to replace the polymer matrix making treatment of spent IX resins cost effective. (author)

  2. Seismic Response Analysis of Spent Nuclear Fuel Metal Storage Cask considering Soil- Structure Interaction Effects

    Energy Technology Data Exchange (ETDEWEB)

    Baeg, Chang-Yeal; Lee, Kyung-Ho; Lee, Dae-Ki [Nuclear Engineering and Technology Institute, Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Jung, In-Su; Song, Won-Tae; Jin, Han-Uk; Kim, Jong-Soo [KONES, Seoul (Korea, Republic of)

    2008-05-15

    Maintaining of the structure safety for the metal storage cask is important to store spent nuclear fuel under a seismic events. Sliding and overturning behavior must be estimated because the metal cask systems are to be installed as free standing structures on reinforced concrete pads. This behavior can cause a serious problem in the integrity of spent nuclear fuel by the impact between neighboring casks. Also, soil condition should be considered since the cask's behavior is strongly affected by the characteristics of the base soil condition. In this study, the seismic response analysis was carried out in order to evaluate the behavior of the metal storage cask under earthquake envelopment considering Soil-Structure Interaction (SSI) effects.

  3. Chemical reactivity testing for the National Spent Nuclear Fuel Program. Revision 2

    International Nuclear Information System (INIS)

    This quality assurance project plan (QAPjP) summarizes requirements used by Lockheed Martin Energy Systems, Incorporated (LMES) Development Division at Y-12 for conducting chemical reactivity testing of Department of Energy (DOE) owned spent nuclear fuel, sponsored by the National Spent Nuclear Fuel Program (NSNFP). The requirements are based on the NSNFP Statement of work PRO-007 (Statement of Work for Laboratory Determination of Uranium Hydride Oxidation Reaction Kinetics.) This QAPjP will utilize the quality assurance program at Y-12, Y60-101PD, Quality Program Description, and existing implementing procedures for the most part in meeting the NSNFP Statement of Work PRO-007 requirements, exceptions will be noted. The project consists of conducting three separate series of related experiments, ''Passivation of Uranium Hydride Powder With Oxygen and Water'', ''Passivation of Uranium Hydride Powder with Surface Characterization'', and ''Electrochemical Measure of Uranium Hydride Corrosion Rate''

  4. Calibration of burnup monitor of spent nuclear fuel installed at Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oeda, Kaoru; Matoba, Masaru; Wakabayashi, Genichiro [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering; Naito, Hirofumi; Hirota, Masanari [Nuclear Fuel Industries Ltd., Tokyo (Japan); Morizaki, Hidetoshi; Kumanomido, Hironori; Natsume, Koichiro [Toshiba Corp., Tokyo (Japan)

    2001-05-01

    The spent nuclear fuel storage pool of Rokkasho reprocessing plant adopts the burnup credit' conception. Spent fuel assemblies are measured every one by one, by burnup monitors, and stored to a storage rack which is designed with specified residual enrichment. For nuclear criticality control, it is necessary for the burnup monitor that the measured value includes a kind of margin, which consists of errors of the monitor. In this paper, we describe the error of the burnup monitors, and the way of taking of the margin. From the result of calibration of the burnup monitor carried out from July through November, 1999, we describe that the way of taking of the margin is validated. And comments about possibility of error reduction are remarked. (author)

  5. Development of U-frame bending system for studying the vibration integrity of spent nuclear fuel

    Science.gov (United States)

    Wang, Hong; Wang, Jy-An John; Tan, Ting; Jiang, Hao; Cox, Thomas S.; Howard, Rob L.; Bevard, Bruce B.; Flanagan, Michelle

    2013-09-01

    A bending fatigue system developed to evaluate the response of spent nuclear fuel rods to vibration loads is presented. A U-frame testing setup is used for imposing bending loads on the fuel rod specimen. The U-frame setup consists of two rigid arms, side connecting plates to the rigid arms, and linkages to a universal testing machine. The test specimen's curvature is obtained through a three-point deflection measurement method. The tests using surrogate specimens with stainless steel cladding revealed increased flexural rigidity under unidirectional cyclic bending, significant effect of cladding-pellets bonding on the response of surrogate rods, and substantial cyclic softening in reverse bending mode. These phenomena may cast light on the expected response of a spent nuclear fuel rod. The developed U-frame system is thus verified and demonstrated to be ready for further pursuit in hot-cell tests.

  6. Compact spent fuel storage at the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    The object of this report is to verify the possibility to increase the available storage of irradiated fuel assemblies, placed in the spent fuel pools of the Atucha I nuclear power plant. There is intends the realization of structural modifications in the storage bracket-suspension beam (single and double) for the upper and lower level of the four spent fuel pools. With these modifications that increase the storage capacity 25%, would arrive until the year 2014, it dates dear for the limit of the commercial operation of nuclear power plant. The increase of the capacity in function of the permissible stress for the supports of the bracket-suspension beam. They should be carried out 5000 re-accommodations of irradiated fuel assemblies. The task would demand approximately 3 years. (author)

  7. Intact and Degraded Component Criticality Calculations of N Reactor Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (keff) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR

  8. SACSESS – the EURATOM FP7 project on actinide separation from spent nuclear fuels

    Directory of Open Access Journals (Sweden)

    Bourg Stéphane

    2015-12-01

    Full Text Available Recycling of actinides by their separation from spent nuclear fuel, followed by transmutation in fast neutron reactors of Generation IV, is considered the most promising strategy for nuclear waste management. Closing the fuel cycle and burning long-lived actinides allows optimizing the use of natural resources and minimizing the long-term hazard of high-level nuclear waste. Moreover, improving the safety and sustainability of nuclear power worldwide. This paper presents the activities striving to meet these challenges, carried out under the Euratom FP7 collaborative project SACSESS (Safety of Actinide Separation Processes. Emphasis is put on the safety issues of fuel reprocessing and waste storage. Two types of actinide separation processes, hydrometallurgical and pyrometallurgical, are considered, as well as related aspects of material studies, process modeling and the radiolytic stability of solvent extraction systems. Education and training of young researchers in nuclear chemistry is of particular importance for further development of this field.

  9. Options for Management of Spent Fuel and Radioactive Waste for Countries Developing New Nuclear Power Programmes

    International Nuclear Information System (INIS)

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world.' One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. Today, numerous countries are considering construction of their first nuclear power plant or the expansion of a small nuclear power programme, and many of these countries have limited experience in managing radioactive waste and spent nuclear fuel. They often have limited information about available technologies and approaches for safe and long term management of radioactive waste and spent nuclear fuel arising from power reactors. The lack of basic know-how and of a credible waste management strategy could present a major challenge or even an obstruction for countries wishing to

  10. Why is a dry storage for spent nuclear fuel waste more appropriate?

    International Nuclear Information System (INIS)

    Nowadays high activity level nuclear waste has been gradually confined in dry shelters, differently than occurred in the previous years. This kind of storages is spreading everywhere. A well evident case is the storage for irradiated fuel elements in commercial nuclear reactors, also known as spent nuclear fuel (SNF). The interim storage, up to 20 years ago, was exclusively carried out through cooling water pools in which the spent fuel has been stored. Such kind of storage was used during the service life of the reactor and sometimes longer. If the nuclear installation had to be decommissioned another storage solution had to be found. At the present time, after a preliminary cooling of the SNF elements inside the water pool, the elements can be stored in dry installations. This kind of storage does not need complex radiation monitoring resulting in favourable construction and maintenance economy of both nuclear storage plants since the volume of the cooling water pools can be smaller. Among several types of existing dry storages, the convection vaults, concrete modules and double purpose casks can be evidenced. These casks can be used as multi-use devices for high activity level materials, i.e. can serve as an interim storage, used for transporting purposes and used as definitive packages for final deposition if so planned. This paper aims to present the advantages of a dry storage in comparison with the wet one (cooling water pools) for high activity level nuclear waste, including the SNF. (author)

  11. Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254

    Energy Technology Data Exchange (ETDEWEB)

    Dilger, Fred [Black Mountain Research, Henderson, NV 81012 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects Carson City, NV 80906 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge Northridge, CA 91330 (United States)

    2012-07-01

    Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National

  12. Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254

    International Nuclear Information System (INIS)

    Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National Laboratories, the 1980's

  13. Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

    OpenAIRE

    Poulson, D.; Durham, J. M.; Guardincerri, E.; Morris, C. L.; Bacon, J. D.; Plaud-Ramos, K.; Morley, D.; Hecht, A.

    2016-01-01

    Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first applicati...

  14. Analysis of Spent Nuclear Fuel Imaging Using Multiple Coulomb Scattering of Cosmic Muons

    OpenAIRE

    Chatzidakis, Stylianos; Choi, Chan K.; Tsoukalas, Lefteri H.

    2016-01-01

    Cosmic ray muons passing through matter lose energy from inelastic collisions with electrons and are deflected from nuclei due to multiple Coulomb scattering. The strong dependence of scattering on atomic number Z and the recent developments on position sensitive muon detectors indicate that multiple Coulomb scattering could be an excellent candidate for spent nuclear fuel imaging. Muons present significant advantages over existing monitoring and imaging techniques and can play a central role...

  15. Stakeholder Preference Mapping—seeking a way forward for the processing of spent nuclear fuel

    OpenAIRE

    Grace McGlynn; Gregg Butler; Alan Pearman

    2015-01-01

    Continuing concern in many countries about the processing of spent nuclear fuel has sparked new interest in how best to make evidence-based decisions about divisive issues. Stakeholder Preference Mapping (SPM), described here, is a way of applying multi-attribute decision analysis to structured dialogue and engagement with stakeholders. It uses the recorded views of stakeholders, supplemented where necessary by direct stakeholder contact, to understand and evidence stakeholder perspectives an...

  16. Direct disposal of spent nuclear fuel. The current status of technology January 1987

    International Nuclear Information System (INIS)

    The Study assesses the degree and status of research and development worldwide on Direct Disposal of Spent Nuclear Fuel. It is limited to technological, radiological and waste management aspects appertaining to Light Water and AGR Reactor Systems and reviews the 'State of the Art' in terms of applicability to the United Kingdom. The report concludes that much technology exists both at National and International level which the UK can apply to the subject of Direct Disposal. (author)

  17. Technical Basis Spent Nuclear Fuel (SNF) Project Radiation and Contamination Trending Program

    Energy Technology Data Exchange (ETDEWEB)

    KURTZ, J.E.

    2000-05-10

    This report documents the technical basis for the Spent Nuclear Fuel (SNF) Program radiation and contamination trending program. The program consists of standardized radiation and contamination surveys of the KE Basin, radiation surveys of the KW basin, and radiation surveys of the Cold Vacuum Drying Facility (CVD) with the associated tracking. This report also discusses the remainder of radiological areas within the SNFP that do not have standardized trending programs and the basis for not having this program in those areas.

  18. STRUCTURAL CALCULATIONS FOR THE CODISPOSAL OF TRIGA SPENT NUCLEAR FUEL IN A WASTE PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    S. Mastilovic

    1999-07-28

    The purpose of this analysis is to determine the structural response of a TRIGA Department of Energy (DOE) spent nuclear fuel (SNF) codisposal canister placed in a 5-Defense High Level Waste (DHLW) waste package (WP) and subjected to a tipover design basis event (DBE) dynamic load; the results will be reported in terms of displacements and stress magnitudes. This activity is associated with the WP design.

  19. 78 FR 40200 - Duke Energy Carolinas, LLC, Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel...

    Science.gov (United States)

    2013-07-03

    ... spent fuel inventory to meet decay heat zoning requirements throughout its scheduled loading campaigns... Nuclear Fuel to the list of approved spent fuel storage casks in 10 CFR 72.214 (59 FR 28496, June 2, 1994 (Proposed Rule); 59 FR 65920, December 22, 1994 (Final Rule)). The exemption proposed to Amendment No. 9...

  20. Treatment and recycling of spent nuclear fuel. Actinide partitioning - Application to waste management

    International Nuclear Information System (INIS)

    subsequent to its in-reactor dwell time, spent fuel still contains large amounts of materials that are recoverable, for value-added energy purposes (uranium, plutonium), together with fission products, and minor actinides, making up the residues from nuclear reactions. The treatment and recycling of spent nuclear fuel, as implemented in France, entail that such materials be chemically partitioned. The development of the process involved, and its deployment on an industrial scale stand as a high achievement of French science, and technology. Treatment and recycling allow both a satisfactory management of nuclear waste to be implemented, and substantial savings, in terms of fissile material. Bolstered of late as it has been, due to spectacularly skyrocketing uranium prices, this strategy is bound to become indispensable, with the advent of the next generation of fast reactors. This Monograph surveys the chemical process used for spent fuel treatment, and its variants, both current, and future. It outlines currently ongoing investigations, setting out the challenges involved, and recent results obtained by CEA. (authors)

  1. Treatment and recycling of spent nuclear fuel. Actinide partitioning - Application to waste management

    Energy Technology Data Exchange (ETDEWEB)

    Abonneau, E.; Baron, P.; Berthon, C.; Berthon, L.; Beziat, A.; Bisel, I.; Bonin, L.; Bosse, E.; Boullis, B.; Broudic, J.C.; Charbonnel, M.C.; Chauvin, N.; Den Auwer, C.; Dinh, B.; Duhamet, J.; Escleine, J.M.; Grandjean, S.; Guilbaud, P.; Guillaneux, D.; Guillaumont, D.; Hill, C.; Lacquement, J.; Masson, M.; Miguirditchian, M.; Moisy, P.; Pelletier, M.; Ravenet, A.; Rostaing, C.; Royet, V.; Ruas, A.; Simoni, E.; Sorel, C.; Vaudano, A.; Venault, L.; Warin, D.; Zaetta, A.; Pradel, P.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Forestier, A.; Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Latge, C.; Limoge, Y.; Madic, C.; Santarini, G.; Seiler, J.M.; Sollogoob, P.; Vernaz, E.; Bazile, F.; Parisot, J.P.; Finot, P.; Roberts, J.F

    2008-07-01

    subsequent to its in-reactor dwell time, spent fuel still contains large amounts of materials that are recoverable, for value-added energy purposes (uranium, plutonium), together with fission products, and minor actinides, making up the residues from nuclear reactions. The treatment and recycling of spent nuclear fuel, as implemented in France, entail that such materials be chemically partitioned. The development of the process involved, and its deployment on an industrial scale stand as a high achievement of French science, and technology. Treatment and recycling allow both a satisfactory management of nuclear waste to be implemented, and substantial savings, in terms of fissile material. Bolstered of late as it has been, due to spectacularly skyrocketing uranium prices, this strategy is bound to become indispensable, with the advent of the next generation of fast reactors. This Monograph surveys the chemical process used for spent fuel treatment, and its variants, both current, and future. It outlines currently ongoing investigations, setting out the challenges involved, and recent results obtained by CEA. (authors)

  2. The Impact of Microbially Influenced Corrosion on Spent Nuclear Fuel and Storage Life

    Energy Technology Data Exchange (ETDEWEB)

    J. H. Wolfram; R. E. Mizia; R. Jex; L. Nelson; K. M. Garcia

    1996-10-01

    A study was performed to evaluate if microbial activity could be considered a threat to spent nuclear fuel integrity. The existing data regarding the impact of microbial influenced corrosion (MIC) on spent nuclear fuel storage does not allow a clear assessment to be made. In order to identify what further data are needed, a literature survey on MIC was accomplished with emphasis on materials used in nuclear fuel fabrication, e.g., A1, 304 SS, and zirconium. In addition, a survey was done at Savannah River, Oak Ridge, Hanford, and the INEL on the condition of their wet storage facilities. The topics discussed were the SNF path forward, the types of fuel, ramifications of damaged fuel, involvement of microbial processes, dry storage scenarios, ability to identify microbial activity, definitions of water quality, and the use of biocides. Information was also obtained at international meetings in the area of biological mediated problems in spent fuel and high level wastes. Topics dis cussed included receiving foreign reactor research fuels into existing pools, synergism between different microbes and other forms of corrosion, and cross contamination.

  3. The Impact of Microbially Influenced Corrosion on Spent Nuclear Fuel and Storage Life

    International Nuclear Information System (INIS)

    A study was performed to evaluate if microbial activity could be considered a threat to spent nuclear fuel integrity. The existing data regarding the impact of microbial influenced corrosion (MIC) on spent nuclear fuel storage does not allow a clear assessment to be made. In order to identify what further data are needed, a literature survey on MIC was accomplished with emphasis on materials used in nuclear fuel fabrication, e.g., A1, 304 SS, and zirconium. In addition, a survey was done at Savannah River, Oak Ridge, Hanford, and the INEL on the condition of their wet storage facilities. The topics discussed were the SNF path forward, the types of fuel, ramifications of damaged fuel, involvement of microbial processes, dry storage scenarios, ability to identify microbial activity, definitions of water quality, and the use of biocides. Information was also obtained at international meetings in the area of biological mediated problems in spent fuel and high level wastes. Topics dis cussed included receiving foreign reactor research fuels into existing pools, synergism between different microbes and other forms of corrosion, and cross contamination

  4. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement

    International Nuclear Information System (INIS)

    This volume contains the following attachments: transportation of Naval spent nuclear fuel; description of Naval spent nuclear receipt and handling at the Expended Core Facility at the Idaho National Engineering Laboratory; comparison of storage in new water pools versus dry container storage; description of storage of Naval spent nuclear fuel at servicing locations; description of receipt, handling, and examination of Naval spent nuclear fuel at alternate DOE facilities; analysis of normal operations and accident conditions; and comparison of the Naval spent nuclear fuel storage environmental assessment and this environmental impact statement

  5. Non destructive assay of nuclear LEU spent fuels for burnup credit application

    International Nuclear Information System (INIS)

    Criticality safety analysis devoted to spent fuel storage and transportation has to be conservative in order to be sure no accident will ever happen. In the spent fuel storage field, the assumption of freshness has been used to achieve the conservative aspect of criticality safety procedures. Nevertheless, after being irradiated in a reactor core, the fuel elements have obviously lost part of their original reactivity. The concept of taking into account this reactivity loss in criticality safety analysis is known as Burnup credit. To be used, Burnup credit involves obtaining evidence of the reactivity loss with a Burnup measurement. Many non destructive assays (NDA) based on neutron as well as on gamma ray emissions are devoted to spent fuel characterization. Heavy nuclei that compose the fuels are modified during irradiation and cooling. Some of them emit neutrons spontaneously and the link to Burnup is a power link. As a result, burn-up determination with passive neutron measurement is extremely accurate. Some gamma emitters also have interesting properties in order to characterize spent fuels but the convenience of the gamma spectrometric methods is very dependent on characteristics of spent fuel. In addition, contrary to the neutron emission, the gamma signal is mostly representative of the peripheral rods of the fuels. Two devices based on neutron methods but combining different NDA methods which have been studied in the past are described in detail: 1. The PYTHON device is a combination of a passive neutron measurement, a collimated total gamma measurement, and an online depletion code. This device, which has been used in several Nuclear Power Plants in western Europe, gives the average Burnup within a 5% uncertainty and also the extremity Burnup, 2. The NAJA device is an automatic device that involves three nuclear methods and an online depletion code. It is designed to cover the whole fuel assembly panel (Active Neutron Interrogation, Passive Neutron

  6. Spent nuclear fuel and high level radioactive waste transportation. White paper

    International Nuclear Information System (INIS)

    The High-Level Radioactive Waste Committee of the Western Interstate Energy Board has been involved in a year-long cooperative project with the US Department of Energy (DOE) to develop an information base on the transportation of spent nuclear fuel and high-level radioactive waste (HLW) so that western states can be constructive and informed participants in the repository program under the Nuclear Waste Policy Act (NWPA). The historical safety record of transportation of HLW and spent fuel is excellent; no release of these radioactive materials has ever occurred during transportation. Projected shipments under the NWPA will, however, greatly exceed current shipments in the US. For example, over the past five years, 119 metric tons of civilian spent fuel have been shipped in this country, while shipments to the first and second repository are each expected to peak at 3000 metric tons per year. The Committee believes that the successful development and operation of a national HLW/spent fuel transportation system can best be accomplished through an open process based on the common sense approach of taking all reasonable measures to minimize public risk and performing whatever actions are reasonably required to promote public acceptance. Therefore, the Committee recommends that the Department of Energy further the goals of the NWPA by developing a Comprehensive Transportation Plan which adopts a systematic, comprehensive, and integrated approach to resolving all spent fuel and HLW transportation issues in a timely manner. The suggested scope of such a plan is discussed in this White paper. Many of the suggested elements of such a plan are similar to those being developed by the Department of energy for inclusion in the Department's Transportation Institutional Plan

  7. Nondestructive assay of Pu in spent fuel using nuclear resonance fluorescence with monochromatic gamma-rays

    International Nuclear Information System (INIS)

    We have proposed a nondestructive assay for Pu-239 in spent fuel assembly using nuclear resonance fluorescence with energy tunable monochromatic gamma-rays generated by Compton scattering of laser photons and high energy electrons. To demonstrate this method, we carried out nuclear experiments using available laser Compton scattering gamma-rays. We measured NRF gamma-rays of Pb-208 concealed in an iron box with a thickness of 15 mm using LCS gamma-rays at National Institute of Advanced Industrial Science and Technology. We also measured NRF gamma-rays of U-238 using LCS gamma-rays at Duke University. (author)

  8. Radiation exposures associated with shipments of foreign research reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    In accordance with the Record of Decision on a Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel (ROD) (DOE, 1996a), the U.S. Department of Energy (DOE) is implementing a 13-year program under which DOE accepts foreign research reactor spent nuclear fuel (SNF) containing uranium that was enriched in the United States. The ROD required that DOE take several steps to ensure low environmental and health impacts resulting from the implementation of the program. These efforts mainly focus on transportation related activities that the analysis of potential environmental impacts in the Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel (EIS) (DOE, 1996b) identified as having the potential for exceeding current radiation protection guidelines. Consequently, DOE issued a Mitigation Action Plan to reduce the likelihood of potential adverse environmental impacts associated with the policy established in the ROD. As shown in the EIS, incident-free radiation exposures to members of the ship's crew, port workers, and ground transportation personnel due to shipments of spent nuclear fuel from foreign research reactors are expected to be below the radiation exposure limit of 100 mrem (1 mSv) per year established to protect the general public. However, the analysis in the EIS demonstrated that port and transportation workers could conceivably receive a cumulative radiation dose above the limit established for the general public if, for example, they are involved in multiple shipments within one year or if the radiation levels outside the casks are at the maximum allowable regulatory limit (10 mrem/hr [0.1 mSv/h] at 2 meters from the surface of the cask). With the program successfully underway, DOE has collected information from the shipments in accordance with the Mitigation Action Plan. The information to date has demonstrated that the analysis in

  9. Choice of method - evaluation of strategies and systems for disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    2010-10-15

    This report deals with the question of how the Swedish spent nuclear fuel is to be disposed of. What are the requirements? What are the alternatives? In the main chapter of the report, an evaluation is made of the KBS-3 method compared with other strategies and systems for final disposal of spent nuclear fuel. An appendix to the report presents in general terms how the KBS-3 method has developed from the end of the 1970s up to today. The report is one of a number of supporting documents for SKB's applications for construction and operation of the final repository for spent nuclear fuel. In parallel with and as a basis for the present report, SKB has prepared the reports Principer, strategier och system foer slutligt omhaendertagande av anvaent kaernbraensle ('Principles, strategies and systems for final disposal of spent nuclear fuel') /Grundfelt 2010a/, Jaemfoerelse mellan KBS-3-metoden och deponering i djupa borrhaal foer slutlig foervaring av anvaent kaernbraensle ('Comparison between the KBS-3 method and deposition in deep boreholes for final disposal of spent nuclear fuel') /Grundfelt 2010b/ and Utvecklingen av KBS-3- metoden. Genomgaang av forskningsprogram, saekerhetsanalyser, myndighetsgranskningar samt SKB:s internationella forskningssamarbete ('Development of the KBS-3 method. Review of research programmes, safety assessments, regulatory reviews and SKB's international research cooperation') /SKB 2010a/. The reports are in Swedish, but contain summaries in English. The first report is an update of the comprehensive account of alternative methods presented by SKB in 2000. The second report presents a comparison between the KBS-3 method and the Deep Boreholes concept, plus a status report on research and development in the area of Deep Boreholes. The last report describes how the KBS-3 method has been developed from the end of the 1970s up to today. It further describes how the method has been further developed and

  10. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    International Nuclear Information System (INIS)

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States [CIS]). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future

  11. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States [CIS]). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  12. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States (CIS)). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  13. Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ludewigt, Bernhard A; Quiter, Brian J.; Ambers, Scott D.

    2011-01-14

    The Next Generation Safeguard Initiative (NGSI) of the U.S Department of Energy is supporting a multi-lab/university collaboration to quantify the plutonium (Pu) mass in spent nuclear fuel (SNF) assemblies and to detect the diversion of pins with non-destructive assay (NDA) methods. The following 14 NDA techniques are being studied: Delayed Neutrons, Differential Die-Away, Differential Die-Away Self-Interrogation, Lead Slowing Down Spectrometer, Neutron Multiplicity, Passive Neutron Albedo Reactivity, Total Neutron (Gross Neutron), X-Ray Fluorescence, {sup 252}Cf Interrogation with Prompt Neutron Detection, Delayed Gamma, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Self-integration Neutron Resonance Densitometry, and Neutron Resonance Transmission Analysis. Understanding and maturity of the techniques vary greatly, ranging from decades old, well-understood methods to new approaches. Nuclear Resonance Fluorescence (NRF) is a technique that had not previously been studied for SNF assay or similar applications. Since NRF generates isotope-specific signals, the promise and appeal of the technique lies in its potential to directly measure the amount of a specific isotope in an SNF assay target. The objectives of this study were to design and model suitable NRF measurement methods, to quantify capabilities and corresponding instrumentation requirements, and to evaluate prospects and the potential of NRF for SNF assay. The main challenge of the technique is to achieve the sensitivity and precision, i.e., to accumulate sufficient counting statistics, required for quantifying the mass of Pu isotopes in SNF assemblies. Systematic errors, considered a lesser problem for a direct measurement and only briefly discussed in this report, need to be evaluated for specific instrument designs in the future. Also, since the technical capability of using NRF to measure Pu in SNF has not been established, this report does not directly address issues such as cost, size

  14. Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    The Next Generation Safeguard Initiative (NGSI) of the U.S Department of Energy is supporting a multi-lab/university collaboration to quantify the plutonium (Pu) mass in spent nuclear fuel (SNF) assemblies and to detect the diversion of pins with non-destructive assay (NDA) methods. The following 14 NDA techniques are being studied: Delayed Neutrons, Differential Die-Away, Differential Die-Away Self-Interrogation, Lead Slowing Down Spectrometer, Neutron Multiplicity, Passive Neutron Albedo Reactivity, Total Neutron (Gross Neutron), X-Ray Fluorescence, 252Cf Interrogation with Prompt Neutron Detection, Delayed Gamma, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Self-integration Neutron Resonance Densitometry, and Neutron Resonance Transmission Analysis. Understanding and maturity of the techniques vary greatly, ranging from decades old, well-understood methods to new approaches. Nuclear Resonance Fluorescence (NRF) is a technique that had not previously been studied for SNF assay or similar applications. Since NRF generates isotope-specific signals, the promise and appeal of the technique lies in its potential to directly measure the amount of a specific isotope in an SNF assay target. The objectives of this study were to design and model suitable NRF measurement methods, to quantify capabilities and corresponding instrumentation requirements, and to evaluate prospects and the potential of NRF for SNF assay. The main challenge of the technique is to achieve the sensitivity and precision, i.e., to accumulate sufficient counting statistics, required for quantifying the mass of Pu isotopes in SNF assemblies. Systematic errors, considered a lesser problem for a direct measurement and only briefly discussed in this report, need to be evaluated for specific instrument designs in the future. Also, since the technical capability of using NRF to measure Pu in SNF has not been established, this report does not directly address issues such as cost, size, development

  15. Accelerator-driven system design concept for disposing of spent nuclear fuels

    International Nuclear Information System (INIS)

    At present, the US SNF (Spent Nuclear Fuel) inventory is growing by about 2,000 metric tonnes (MT) per year from the current operating nuclear power plants to reach about 70,000 MT by 2015. This SNF inventory contains about 1% transuranics (700 MT), which has about 115 MT of minor actinides. Accelerator-driven systems utilising proton accelerators with neutron spallation targets and subcritical blankets can be utilised for transmuting these transuranics, simultaneously generating carbon free energy, and significantly reducing the capacity of the required geological repository storage facility for the spent nuclear fuels. A fraction of the SNF plutonium can be used as a MOX fuel in the current/future thermal power reactors and as a starting fuel for future fast power reactors. The uranium of the spent nuclear fuel can be recycled for use in future nuclear power plants. This paper shows that only four to five accelerator-driven systems operating for less than 33 full power years can dispose of the US SNF inventory expected by 2015. In addition, a significant fraction of the long-lived fission products will be transmuted at the same time. Each system consists of a proton accelerator with a neutron spallation target and a subcritical assembly. The accelerator beam parameters are 1 GeV protons and 25 MW beam power, which produce 3 GWt in the subcritical assembly. A liquid metal (lead or lead-bismuth eutectic) spallation target is selected because of design advantages. This target is located at the centre of the subcritical assembly to maximise the utilisation of spallation neutrons. Because of the high power density in the target material, the target has its own coolant loop, which is independent of the subcritical assembly coolant loop. Mobile fuel forms with transuranic materials without uranium are considered in this work with liquid lead or lead-bismuth eutectic as fuel carrier

  16. Truck and rail charges for shipping spent fuel and nuclear waste

    International Nuclear Information System (INIS)

    The Pacific Northwest Laboratory developed techniques for calculating estimates of nuclear-waste shipping costs and compiled a listing of representative data that facilitate incorporation of reference shipping costs into varius logistics analyses. The formulas that were developed can be used to estimate costs that will be incurred for shipping spent fuel or nuclear waste by either legal-weight truck or general-freight rail. The basic data for this study were obtained from tariffs of a truck carrier licensed to serve the 48 contiguous states and from various rail freight tariff guides. Also, current transportation regulations as issued by the US Department of Transportation and the Nuclear Regulatory Commission were investigated. The costs that will be incurred for shipping spent fuel and/or nuclear waste, as addressed by the tariff guides, are based on a complex set of conditions involving the shipment origin, route, destination, weight, size, and volume and the frequency of shipments, existing competition, and the length of contracts. While the complexity of these conditions is an important factor in arriving at a ''correct'' cost, deregulation of the transportation industry means that costs are much more subject to negotiation and, thus, the actual fee that will be charged will not be determined until a shipping contract is actually signed. This study is designed to provide the baseline data necessary for making comparisons of the estimated costs of shipping spent fuel and/or nuclear wastes by truck and rail transportation modes. The scope of the work presented in this document is limited to the costs incurred for shipping, and does not include packaging, cask purchase/lease costs, or local fees placed on shipments of radioactive materials

  17. Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

    Science.gov (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment. PMID:16381764

  18. OECD/NEA expert group on assay data of spent nuclear fuel

    International Nuclear Information System (INIS)

    In the area of criticality safety, management of spent nuclear fuel is a key issue for many NEA member countries. The importance of measured isotopic assay data from Post-Irradiation Examination (PIE) experiments to validate computer code predictions of spent fuel composition used in safety-related studies has long been recognized by members of the OECD/NEA/NSC/WPNCS (Working Party on Nuclear Criticality Safety). These data are particularly important in criticality analyses related to any application of burnup credit as well as to evaluation of criticality and safety in geologic repositories and fuel cycle applications such as reprocessing. Under the auspices of the WPNCS, an Expert Group on assay data has been formed to share best-practice radiochemical analysis methods, computational analysis procedures and data needs, and isotopic validation data. Through member country collaboration, the database of publicly available spent fuel measurements is being revised and expanded to include more recent measurements, with findings to be documented in a state-of-the-art report. (author)

  19. Spent Nuclear Fuel Project FY 1996 Multi-Year Program Plan WBS No. 1.4.1, Revision 1

    International Nuclear Information System (INIS)

    This document describes the Spent Nuclear Fuel (SNF) Project portion of the Hanford Strategic Plan for the Hanford Reservation in Richland, Washington. The SNF Project was established to evaluate and integrate the urgent risks associated with N-reactor fuel currently stored at the Hanford site in the K Basins, and to manage the transfer and disposition of other spent nuclear fuels currently stored on the Hanford site. An evaluation of alternatives for the expedited removal of spent fuels from the K Basin area was performed. Based on this study, a Recommended Path Forward for the K Basins was developed and proposed to the U.S. DOE

  20. Developing the Technical Data Supporting Licensing of Very Long Term Dry Storage of Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    At the last meeting of this series in 2006, license renewal and long-term storage were identified as topics that needed further consideration in the near future. Since that time three independent actions took place: 1) In June 2009 the Nuclear Waste Technical Review Board, conducted a public meeting to review the information available to provide a sound regulatory and technical basis for the safe and secure long-term storage of spent nuclear fuel (SNF), 2) On February 18, 2010, the Commissioners of the United States Nuclear Regulatory Commission (NRC) directed the NRC staff to undertake a thorough review of the regulatory programs for spent fuel storage and transportation, including their adequacy for ensuring safe and secure storage and transportation of spent fuel for extended periods beyond the 120-year timeframe considered up to this point. [1], and 3) On March 3, 2010, the US Department of Energy (DOE) motioned to withdraw its pending license application for Yucca Mountain as a permanent site for disposal of SNF. This action could lead to a need for very long-term dry storage (VLTDS) of spent nuclear fuel (SNF). All of the above suggest a need for better technical data to support the safe and secure storage of SNF for a long time and to be able to remain in a transportable condition without repackaging. To date, the staff of the Spent Fuel Storage and Transportation Division (SFST) of the NRC has observed no data or operating experience that indicates that SNF cannot be safely and securely stored for a very long term and remain in a transportable configuration in large casks under the proper storage conditions with an appropriate aging management plan. However, the staff currently believes that additional data is necessary to confirm and to demonstrate that VLTDS and subsequent transportation, without repackaging, can be safely accomplished. To address this potential need, the SFST staff is developing a plan for Commission approval to address both the

  1. SPENT NUCLEAR FUEL STORAGE BASIN WATER CHEMISTRY: ELECTROCHEMICAL EVALUATION OF ALUMINUM CORROSION

    Energy Technology Data Exchange (ETDEWEB)

    Hathcock, D

    2007-10-30

    The factors affecting the optimal water chemistry of the Savannah River Site spent fuel storage basin must be determines in order to optimize facility efficiency, minimize fuel corrosion, and reduce overall environmental impact from long term spent nuclear fuel storage at the Savannah River Site. The Savannah River National Laboratory is using statistically designed experiments to study the effects of NO{sub 3}{sup -}, SO{sub 4}{sup 2-}, and Cl{sup -} concentrations on alloys commonly used not only as fuel cladding, but also as rack construction materials The results of cyclic polarization pitting and corrosion experiments on samples of Al 6061 and 1100 alloys will be used to construct a predictive model of the basin corrosion and its dependence on the species in the basin. The basin chemistry model and corrosion will be discussed in terms of optimized water chemistry envelope and minimization of cladding corrosion.

  2. Safety case for the disposal of spent nuclear fuel at Olkiluoto - Synthesis 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-12-15

    TURVA-2012 is Posiva's safety case in support of the Preliminary Safety Analysis Report (PSAR 2012) and application for a construction licence for a spent nuclear fuel repository. Consistent with the Government Decisions-in- Principle, this foresees a repository developed in bedrock at the Olkiluoto site according to the KBS-3 method, designed to accept spent nuclear fuel from the lifetime operations of the Olkiluoto and Loviisa reactors. Synthesis 2012 presents a synthesis of Posiva Oy's Safety Case 'TURVA-2012' portfolio. It summarises the design basis for the repository at the Olkiluoto site, the assessment methodology and key results of performance and safety assessments. It brings together all the lines of argument for safety, evaluation of compliance with the regulatory requirements, and statement of confidence in long-term safety and Posiva's safety analyses. The TURVA-2012 safety case demonstrates that the proposed repository design provides a safe solution for the disposal of spent nuclear fuel, and that the performance and safety assessments are fully consistent with all the legal and regulatory requirements related to long-term safety as set out in Government Decree 736/2008 and in guidance from the nuclear regulator - the STUK. Moreover, Posiva considers that the level of confidence in the demonstration of safety is appropriate and sufficient to submit the construction licence application to the authorities. The assessment of long-term safety includes uncertainties, but these do not affect the basic conclusions on the long-term safety of the repository. (orig.)

  3. THE ECONOMICS OF REPROCESSING vs DIRECT DISPOSAL OF SPENT NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Matthew Bunn; Steve Fetter; John P. Holdren; Bob van der Zwaan

    2003-07-01

    This report assesses the economics of reprocessing versus direct disposal of spent nuclear fuel. The breakeven uranium price at which reprocessing spent nuclear fuel from existing light-water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is assessed, using central estimates of the costs of different elements of the nuclear fuel cycle (and other fuel cycle input parameters), for a wide range of range of potential reprocessing prices. Sensitivity analysis is performed, showing that the conclusions reached are robust across a wide range of input parameters. The contribution of direct disposal or reprocessing and recycling to electricity cost is also assessed. The choice of particular central estimates and ranges for the input parameters of the fuel cycle model is justified through a review of the relevant literature. The impact of different fuel cycle approaches on the volume needed for geologic repositories is briefly discussed, as are the issues surrounding the possibility of performing separations and transmutation on spent nuclear fuel to reduce the need for additional repositories. A similar analysis is then performed of the breakeven uranium price at which deploying fast neutron breeder reactors would become competitive compared with a once-through fuel cycle in LWRs, for a range of possible differences in capital cost between LWRs and fast neutron reactors. Sensitivity analysis is again provided, as are an analysis of the contribution to electricity cost, and a justification of the choices of central estimates and ranges for the input parameters. The equations used in the economic model are derived and explained in an appendix. Another appendix assesses the quantities of uranium likely to be recoverable worldwide in the future at a range of different possible future prices.

  4. Spent-Fuel Test - Climax: An evaluation of the technical feasibility of geologic storage of spent nuclear fuel in granite: Executive summary of final results

    International Nuclear Information System (INIS)

    This summary volume outlines results that are covered in more detail in the final report of the Spent-Fuel Test - Climate project. The project was conducted between 1978 and 1983 in the granitic Climax stock at the Nevada Test Site. Results indicate that spent fuel can be safely stored for periods of years in this host medium and that nuclear waste so emplaced can be safely retrieved. We also evaluated the effects of heat and radiation (alone and in combination) on emplacement canisters and the surrounding rock mass. Storage of the spent-fuel affected the surrounding rock mass in measurable ways, but did not threaten the stability or safety of the facility at any time

  5. Evaluation of the recycling costs, as a disposal form of the spent nuclear fuel

    International Nuclear Information System (INIS)

    At the moment there are 2 BWR reactors operating in the Nuclear Power station of Laguna Verde in Mexico. At the end of the programmed life of the reactors (40 years) its will have completed 26 operation cycles, with will have 6712 spent fuel assemblies will be in the pools of the power station. Up to now, the decision on the destination of the high level wastes (spent nuclear fuel) it has not been determined in Mexico, the same as in other countries, adopting a politics of 'to wait to see that it happens in the world', in this respect, in the world two practical alternatives exist, one is to store the fuel in repositories designed for that end, another is reprocess the fuel to recycle the plutonium contained in it, both solutions have their particular technical and economic problematic. In this work it is evaluated from the economic point of view the feasibility of having the spent fuel, using the one recycled fuel, for that which thinks about a consistent scenario of a BWR reactor in which the fuel discharged in each operation cycle is reprocessed and its are built fuel assemblies of the MOX type to replace partly to the conventional fuel. This scenario shows an alternative to the indefinite storage of the high level radioactive waste. The found results when comparing from the economic point of view both options, show that the one recycled, even with the current costs of the uranium it is of the order of 7% more expensive that the option of storing the fuel in repositories constructed for that purpose. However the volumes of spent fuel decrease in 66%. (Author)

  6. Modeling of molecular and particulate transport in dry spent nuclear fuel canisters

    Science.gov (United States)

    Casella, Andrew M.

    2007-09-01

    The transportation and storage of spent nuclear fuel is one of the prominent issues facing the commercial nuclear industry today, as there is still no general consensus regarding the near- and long-term strategy for managing the back-end of the nuclear fuel cycle. The debate continues over whether the fuel cycle should remain open, in which case spent fuel will be stored at on-site reactor facilities, interim facilities, or a geologic repository; or if the fuel cycle should be closed, in which case spent fuel will be recycled. Currently, commercial spent nuclear fuel is stored at on-site reactor facilities either in pools or in dry storage containers. Increasingly, spent fuel is being moved to dry storage containers due to decreased costs relative to pools. As the number of dry spent fuel containers increases and the roles they play in the nuclear fuel cycle increase, more regulations will be enacted to ensure that they function properly. Accordingly, they will have to be carefully analyzed for normal conditions, as well as any off-normal conditions of concern. This thesis addresses the phenomena associated with one such concern; the formation of a microscopic through-wall breach in a dry storage container. Particular emphasis is placed on the depressurization of the canister, release of radioactivity, and plugging of the breach due to deposition of suspended particulates. The depressurization of a dry storage container upon the formation of a breach depends on the temperature and quantity of the fill gas, the pressure differential across the breach, and the size of the breach. The first model constructed in this thesis is capable of determining the depressurization time for a breached container as long as the associated parameters just identified allow for laminar flow through the breach. The parameters can be manipulated to quantitatively determine their effect on depressurization. This model is expanded to account for the presence of suspended particles. If

  7. Nuclear Dynamics Consequence Analysis (NDCA) for the Disposal of Spent Nuclear Fuel in an Underground Geologic Repository - Volume 3: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, L.L.; Wilson, J.R. (INEEL); Sanchez, L.C.; Aguilar, R.; Trellue, H.R.; Cochrane, K. (SNL); Rath, J.S. (New Mexico Engineering Research Institute)

    1998-10-01

    The United States Department of Energy Office of Environmental Management's (DOE/EM's) National Spent Nuclear Fuel Program (NSNFP), through a collaboration between Sandia National Laboratories (SNL) and Idaho National Engineering and Environmental Laboratory (INEEL), is conducting a systematic Nuclear Dynamics Consequence Analysis (NDCA) of the disposal of SNFs in an underground geologic repository sited in unsaturated tuff. This analysis is intended to provide interim guidance to the DOE for the management of the SNF while they prepare for final compliance evaluation. This report presents results from a Nuclear Dynamics Consequence Analysis (NDCA) that examined the potential consequences and risks of criticality during the long-term disposal of spent nuclear fuel owned by DOE-EM. This analysis investigated the potential of post-closure criticality, the consequences of a criticality excursion, and the probability frequency for post-closure criticality. The results of the NDCA are intended to provide the DOE-EM with a technical basis for measuring risk which can be used for screening arguments to eliminate post-closure criticality FEPs (features, events and processes) from consideration in the compliance assessment because of either low probability or low consequences. This report is composed of an executive summary (Volume 1), the methodology and results of the NDCA (Volume 2), and the applicable appendices (Volume 3).

  8. REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, T.; Beals, D.; Sternat, M.

    2011-07-18

    Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical

  9. Reactor Physics Modeling Of Spent Research Reactor Fuel For Technical Nuclear Forensics

    International Nuclear Information System (INIS)

    Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to ∼93% 235U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The

  10. Spent nuclear fuel management system in the Russian Federation, SNF stockpile and plans for reprocessing

    International Nuclear Information System (INIS)

    A strategic policy of the development of nuclear industry in the Russian Federation is the closure of nuclear fuel cycle, which must result in a more efficient use of natural uranium recourses and artificial fissile materials generated in nuclear reactors (RepU, plutonium, etc.), ensuring minimization of waste from fuel reprocessing and the wastes for final disposal. The main idea is to provide the reprocessing service for the major part of irradiated fuel including legacy at the radiochemical facilities both existing and under construction. Relations in the field of spent fuel management and of radioactive waste management are regulated on the basis of the Constitution of the Russian Federation as the Basic Law having the supreme legal force and direct action, federal laws of the Russian Federation, legal acts by the President of the Russian Federation and the Government of the Russian Federation, federal norms and rules for the activity in the field of atomic energy use, regulations of the state regulatory authorities as well as relevant national standards. The basic laws which directly regulate relations in the field of safety include Federal Laws: 'On the Use of Atomic Energy', 'On the Radiation Safety of the Public', 'On the Environmental Protection', 'On the Sanitary and Epidemiologic Safety of the Public'. The efficient Spent Fuel Management in the Russian Federation is guided by the following criteria: Nuclear and radiological Safety; RadWastes Minimization; Economical use / disposition of the secondary nuclear materials taking into account quality of the materials; Compliance with the IAEA Convention; Cost optimization. This approach has been supported by the major events of 2006, among which: 1) 25 January 2006 - the Russian Federation President's Initiative with the offer to develop the Global Nuclear Infrastructure with International Centres for the Nuclear Fuel Cycle Service. 2) 6 October 2006 - Approval by the Russian Government of the Federal Target

  11. Radiological impacts of spent nuclear fuel management options. A comparative study

    International Nuclear Information System (INIS)

    Given its potential significance for public health and the environment, the impact of radioactive releases during important steps of nuclear energy production must be considered when selecting among different fuel cycles. With this in mind, the OECD Nuclear Energy Agency (NEA) has undertaken a comparative study to the radiological impacts of two main fuel cycle options : one with and one without reprocessing of spent nuclear fuel. The study compares the respective impacts of the two options based on generic models and assumptions as well as actual data. It concludes that the difference between them is not significant. A wealth of recent data assembled and evaluated by an international expert team is provided in annex. (authors)

  12. Integrated scheme of long-term for spent fuel management of power nuclear reactors

    International Nuclear Information System (INIS)

    After of irradiation of the nuclear fuel in the reactor core, is necessary to store it for their cooling in the fuel pools of the reactor. This is the first step in a processes series before the fuel can reach its final destination. Until now there are two options that are most commonly accepted for the end of the nuclear fuel cycle, one is the open nuclear fuel cycle, requiring a deep geological repository for the fuel final disposal. The other option is the fuel reprocessing to extract the plutonium and uranium as valuable materials that remaining in the spent fuel. In this study the alternatives for the final part of the fuel cycle, which involves the recycling of plutonium and the minor actinides in the same reactor that generated them are shown. The results shown that this is possible in a thermal reactor and that there are significant reductions in actinides if they are recycled into reactor fuel. (Author)

  13. The X/Q values unit doses for spent nuclear fuel project

    International Nuclear Information System (INIS)

    The purpose of this document is to provide a single referenceable document that provides the X/Qs for all the facilities in the spent nuclear fuel projects, and includes the bases for the X/Q calculations. The X/Q values for the nuclear fuel projects were calculated over the past several years. The values currently used in the nuclear fuel project were documented in letter reports and as attached to various PSEs and Safety Analysis documents. Therefore, there is a need to consolidate these documents or reports into a single referenceable document. The final document includes the X/Qs for KE and KW Basins, the cold vacuum drying facility, and the canister storage building

  14. Available reprocessing and recycling services for research reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    International activities in the back-end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. These programmes will soon have achieved their goals and the SNF take-back programmes will cease. However, the needs of the nuclear community dictate that the majority of the research reactors continue to operate using low enriched uranium (LEU) fuel in order to meet the varied mission objectives. As a result, inventories of LEU SNF will continue to be created and the back-end solution of RR SNF remains a critical issue. In view of this fact, the IAEA, based on the experience gained during the decade of international cooperation in supporting the objectives of the HEU take-back programmes, will draw up a report presenting available reprocessing and recycling services for research reactor spent nuclear fuel. This paper gives an overview of the guiding document which will address all aspects of Reprocessing and Recycling Services for RR SNF, including an overview of solutions, decision making support, service suppliers, conditions (prerequisites, options, etc.), services offered by the managerial and logistics support providers with a focus on available transport packages and applicable transport modes.

  15. Siting of a deep geological repository for spent nuclear fuel in Sweden

    International Nuclear Information System (INIS)

    The siting and construction of a deep geological repository for spent nuclear fuel and other long-lived waste is one of the major remaining tasks within the Swedish Nuclear Waste Management program. In Sweden there is more than 15 years of experience of research and development work concerned with the safe disposal of spent nuclear fuel. This experience encompasses, for example: geological site investigations at several study sites; disposal system concept development and evaluation of alternatives; development and testing of site characterization methods, engineered barrier performance and radionuclide migration experimentation/modeling within the International Stripa Project; major safety assessment studies (KBS-3 and SKB-91); ongoing construction and R and D-work at the Aespoe Hard Rock Laboratory. Given the level of knowledge and experience that has been obtained over these years, the time has now come to systematically proceed towards an actual implementation of a safe system for disposal. A program for a stepwise implementation of a disposal system was presented by SKB to the Swedish government in September 1992. This paper describes in particular the approach to and plans for siting. The siting process is presented and the planned program for site investigations is discussed. Siting of the repository will be done in stages and the work will go on during the 1990's and a few years into the next century before a license to construct a repository can be expected

  16. Safety case for the disposal of spent nuclear fuel at Olkiluoto. Complementary considerations 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-12-15

    Complementary Considerations sits within Posiva Oy's Safety Case 'TURVA-2012' report portfolio and has the objective of enhancing confidence in the outcomes of the safety assessment for a spent nuclear fuel repository to be constructed at Olkiluoto, Finland. The main emphasis in this report is on the evidence and understanding that can be gained from observations at the site, including its regional geological environment, and from natural and anthropogenic analogues for the repository, its components and the processes that affect safety. In particular, the report addresses diverse and less quantifiable types of evidence and arguments that are enclosed to enhance confidence in the outcome of the safety assessment. These complementary considerations have been described as evaluations, evidence and qualitative supporting arguments that lie outside the scope of the other reports of the quantitative safety assessment. The experience with natural analogues for the long-term durability of the materials involved and the extent of processes provides high confidence in our understanding of the disposal system and its evolution. For each engineered barrier and key process, there is increasing analogue evidence to support the conceptual models and parameters. Regarding the suitability of the Olkiluoto site to host a spent fuel repository, a number of factors have been identified that indicate the suitability of crystalline host rock in general, and that of the Olkiluoto site in particular. The report also provides radiation background information for the use of complementary indicators, which aid in putting the results of the safety analysis presented in Assessment of Radionuclide Release Scenarios for the Repository System and Biosphere Assessment in a broader perspective to show that the radiation originating from a spent nuclear fuel repository remains in most cases much below natural background radiation or that caused by non-nuclear industries. (orig.)

  17. Technical and economic aspects of retrievability in spent nuclear fuel disposal. A literature survey

    International Nuclear Information System (INIS)

    This literature research collects cost and feasibility aspects of retrievability of spent nuclear fuel disposal. The Finnish plan for the final disposal of spent nuclear fuel includes some features favouring retrievability, like modularity of disposal plan, good corrosion resistance of copper canisters and long estimated lifetime of disposal shafts and tunnels. Possible political and technical arguments for retrievability are collected and the possible actions taken after retrieving (opening of copper canisters, transportations and interim storage) are looked into. In the Finnish disposal plan retrieving is possible in each stage of repository development. Technical feasibility of retrieving is presented before sealing of a disposal hole, after filling and sealing of a disposal tunnel and after final closure of the repository. Many features of the disposal concept (hydrogeological conditions, radioactive radiation and heat generation of the spent fuel, and properties of bentonite) affect the feasibility and cost of retrieving. Questions dealing with nuclear material safeguards, monitoring and continuity of knowledge should be taken into account when retrievability is being considered. Different ways used for expressing the cost of retrieving, are currency unit, delay of the scheduled disposal project and percentages of the actual disposal cost. Estimated costs for two different cases of retrievability, immediate closure and postponed closure of repository, are presented. Cost estimates, made in Finland, are quite rough, but the Swedish cost calculation example of immediate closure case is interesting because of similarity of Finnish and Swedish disposal concepts. There are also examples of cost, based on research and development work enhancing retrievability. It is noticed that there are only few exact cost estimations relating to retrievability. The reasons might be the improbability of actual retrieving and the fact that retrieving, if materialised, will take

  18. Safety case for the disposal of spent nuclear fuel at Olkiluoto. Complementary considerations 2012

    International Nuclear Information System (INIS)

    Complementary Considerations sits within Posiva Oy's Safety Case 'TURVA-2012' report portfolio and has the objective of enhancing confidence in the outcomes of the safety assessment for a spent nuclear fuel repository to be constructed at Olkiluoto, Finland. The main emphasis in this report is on the evidence and understanding that can be gained from observations at the site, including its regional geological environment, and from natural and anthropogenic analogues for the repository, its components and the processes that affect safety. In particular, the report addresses diverse and less quantifiable types of evidence and arguments that are enclosed to enhance confidence in the outcome of the safety assessment. These complementary considerations have been described as evaluations, evidence and qualitative supporting arguments that lie outside the scope of the other reports of the quantitative safety assessment. The experience with natural analogues for the long-term durability of the materials involved and the extent of processes provides high confidence in our understanding of the disposal system and its evolution. For each engineered barrier and key process, there is increasing analogue evidence to support the conceptual models and parameters. Regarding the suitability of the Olkiluoto site to host a spent fuel repository, a number of factors have been identified that indicate the suitability of crystalline host rock in general, and that of the Olkiluoto site in particular. The report also provides radiation background information for the use of complementary indicators, which aid in putting the results of the safety analysis presented in Assessment of Radionuclide Release Scenarios for the Repository System and Biosphere Assessment in a broader perspective to show that the radiation originating from a spent nuclear fuel repository remains in most cases much below natural background radiation or that caused by non-nuclear industries. (orig.)

  19. Review study 1995. Localization of the repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    This report gives an overview of the studies performed by SKB pertinent to selection of a site for the Swedish repository for spent nuclear fuels, and is written for both experts in the various fields involved, decision-makers and the interested general public. The review can not comprise all detailed factors necessary for deciding the localization, but deals mainly with conditions on the land surface and can point out areas which are not well suited or less interesting as a site. It also treats several scientific, technical and social bases in different parts of the country. 120 refs, 53 figs

  20. Technical Basis Spent Nuclear Fuel (SNF) Project Radiation and Contamination Trending Program

    Energy Technology Data Exchange (ETDEWEB)

    ELGIN, J.C.

    2000-10-02

    This report documents the technical basis for the Spent Nuclear Fuel (SNF) Program radiation and contamination trending program. The program consists of standardized radiation and contamination surveys of the KE Basin, radiation surveys of the KW basin, radiation surveys of the Cold Vacuum Drying Facility (CVD), and radiation surveys of the Canister Storage Building (CSB) with the associated tracking. This report also discusses the remainder of radiological areas within the SNFP that do not have standardized trending programs and the basis for not having this program in those areas.

  1. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2000-11-03

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. The purpose of this revision is to document completion of verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those

  2. Tribal and stakeholder communication and participation strategy for the Spent Nuclear Fuel Project

    International Nuclear Information System (INIS)

    This document outlines a plan to ensure the effective involvement of the Hanford stakeholders and Tribal Governments in Spent Nuclear Fuel (SNF) Project issues and decisions. Stakeholders are defined as the public, news media, regulators, employees, Hanford Advisory Board and members of local, state, and federal governments. Experience at Hanford has shown that early and continued involvement of all interested parties in decision making is absolutely essential for fostering project success. Failure to recognize the importance of this interaction has resulted in significant cost in terms of time and money for several site programs

  3. Integrated data base report--1995: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    The information in this report summarizes the U.S. Department of Energy (DOE) data base for inventories, projections, and characteristics of domestic spent nuclear fuel and radioactive waste. This report is updated annually to keep abreast of continual waste inventory and projection changes in both the government and commercial sectors. Baseline information is provided for DOE program planning purposes and to support DOE program decisions. Although the primary purpose of this document is to provide background information for program planning within the DOE community, it has also been found useful by state and local governments, the academic community, and some private citizens

  4. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  5. Comments on the report 'Environmental Impact Assessments for the final disposal of spent nuclear fuel etc'

    International Nuclear Information System (INIS)

    A working group representing several Swedish governmental authorities provides a detailed description of the legislation currently in effect, which can be applied in the licensing of facilities for the disposal of spent nuclear fuel. The report also describes the current responsibilities among the regulatory authorities. The author's main criticism is that the group has restricted itself within the old framework of present Swedish law and regulatory authority structure, rather than to focus on the EIA concept and in a later stage see whether anything of the present structures can be saved and put into a modern EIA concept

  6. Evaluation of Equivalent Dose Rate of Interim Dry Storage Casks Loaded with Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Equivalent dose rate calculations of the CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks were performed using SCALE 4.3 computer codes system. These casks are planned for an interim storage of spent nuclear fuel at Ignalina NPP. The dose rate calculations were made on the sidelong, upper and lower surface of the cask and at the certain distance. Results show that dose rate values on the surface of the cask are much less then permissible value 1000 μSv/h when average burnup of fuel assembly is 20 GWd/tU. (author)

  7. Apparatus and method for reprocessing and separating spent nuclear fuels. [Patent application

    Science.gov (United States)

    Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.

    1982-01-19

    Spent nuclear fuels, including actinide fuels, volatile and non-volatile fission products, are reprocessed and separated in a molten metal solvent housed in the reaction region of a separation vessel which includes a reflux region positioned above the molten tin solvent. The reflux region minimizes loss of evaporated solvent during the separation of the actinide fuels from the volatile fission products. Additionally, inclusion of the reflux region permits the separation of the more volatile fission products (noncondensable) from the less volatile ones (condensable).

  8. Actinides in metallic waste from electrometallurgical treatment of spent nuclear fuel

    Science.gov (United States)

    Janney, D. E.; Keiser, D. D.

    2003-09-01

    Argonne National Laboratory has developed a pyroprocessing-based technique for conditioning spent sodium-bonded nuclear-reactor fuel in preparation for long-term disposal. The technique produces a metallic waste form whose nominal composition is stainless steel with 15 wt.% Zr (SS-15Zr), up to ˜ 11 wt.% actinide elements (primarily uranium), and a few percent metallic fission products. Actual and simulated waste forms show similar eutectic microstructures with approximately equal proportions of iron solid solution phases and Fe-Zr intermetallics. This article reports on an analysis of simulated waste forms containing uranium, neptunium, and plutonium.

  9. A structural analysis on the KN-12 spent nuclear fuel transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dew Hey [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Lee, Jae Hyung; Na, Jae Yun [Chungnam National Univ., Taejon (Korea, Republic of)

    2001-08-15

    In this study, safety of the spent nuclear fuel cask KN-12 which is developed in 2000 is evaluated for hypothetical accidents conditions such as free drop, puncture, fire accident and water immersion. Finite element code ABAQUS/Explicit is used to compare with safety analysis report of the GNB in which analysis is performed with LS-DYNA3D for hypothetical accident conditions. Through this study, the safety of KN-12 is evaluated by comprehensive structural analysis. The capability and technological advancement of Korean community on the analysis and structural assessment of the cask will be improved. Also people's anxiety about radioactive dangers will be eliminated.

  10. Hazard Evaluation for Storage of Spent Nuclear Fuel (SNF) Sludge at the Solid Waste Treatment Facility

    International Nuclear Information System (INIS)

    This document describes the hazards and operability (HAZOP) analysis performed for the activities associated with the storage of Spent Nuclear Fuel (SNF) sludge at the T Plant. The safety authorization basis for the T Plant for SNF sludge storage is defined as those aspects of facility design and operational requirements relied upon by DOE to authorize operations. The safety authorization basis for the current T Plant activities is described in HNF-SD-WM-ISB-006, Rev 2, ''Interim Safety Basis for Solid Waste Facilities (T Plant) (ISB)'' and associated amendments and engineering change notices. In addition the authorization basis includes other document as summarized in Section 1.3

  11. Preliminary design specification for Department of Energy standardized spent nuclear fuel canisters. Volume 2: Rationale document

    International Nuclear Information System (INIS)

    This document (Volume 2) is a companion document to a preliminary design specification for the design of canisters to be used during the handling, storage, transportation, and repository disposal of Department of Energy (DOE) spent nuclear fuel (SNF). This document contains no procurement information, such as the number of canisters to be fabricated, explicit timeframes for deliverables, etc. However, this rationale document does provide background information and design philosophy in order to help engineers better understand the established design criteria (contained in Volume 1 respectively) necessary to correctly design and fabricate these DOE SNF canisters

  12. DOE Spent Nuclear Fuel Group in Support of Criticality, DBE, TSPA-LA

    Energy Technology Data Exchange (ETDEWEB)

    Henry Loo

    2000-05-01

    This report presents the basis for grouping the over 250 Department of Energy (DOE) spent nuclear fuel (SNF) types in support of analyses for final repository disposal. For each of the required analyses, the parameters needed in conducting the analyses were identified and reviewed. The grouping proposed for the three types of analyses (criticality, design basis events, and total system performance assessment) are based on the similarities of DOE SNF as a function of these parameters. As necessary, further justifications are provided to further reduce the DOE SNF grouping in support of the Office of Civilian Radioactive Waste Management System’s preclosure and postclosure safety cases.

  13. Retrievability of high level waste and spent nuclear fuel. Proceedings of an international seminar

    International Nuclear Information System (INIS)

    The possibility of retrieving spent nuclear fuel or reprocessing high-level radioactive wastes placed in geological repositories is an issue that has attracted increased attention during the past few years, not only among technical experts but also among politicians at different levels, environmental organisations and other interested representatives of the public. This publication contains the presented invited papers, an edited record of the discussions and some concluding remarks. The seminar addressed a wide range of aspects of retrievability including technical options; public acceptance; ethical aspects; long term monitoring and cost considerations; safety and regulatory aspects. Each of the presented papers was indexed separately

  14. Licensing process characteristics of Small Modular Reactors and spent nuclear fuel repository

    Energy Technology Data Exchange (ETDEWEB)

    Söderholm, Kristiina, E-mail: kristiina.soderholm@fortum.com [Fortum Power (Finland); Tuunanen, Jari, E-mail: jari.tuunanen@fortum.com [Fortum Power (Finland); Amaba, Ben, E-mail: baamaba@us.ibm.com [IBM Complex Systems (United States); Bergqvist, Sofia, E-mail: sofia.bergqvist@se.ibm.com [IBM Rational Software (Sweden); Lusardi, Paul, E-mail: plusardi@nuscalepower.com [NuScale Power (United States)

    2014-09-15

    Highlights: • We examine the licensing process challenges of modular nuclear facilities. • We compare the features of Small Modular Reactors and spent nuclear fuel repository. • We present the need of nuclear licensing simplification. • Part of the licensing is proposed to be internationally applicable. • Systems engineering and requirements engineering benefits are presented. - Abstract: This paper aims to increase the understanding of the licensing processes characteristics of Small Modular Reactors (SMR) compared with licensing of spent nuclear fuel repository. The basis of the SMR licensing process development lies in licensing processes used in Finland, France, the UK, Canada and the USA. These countries have been selected for this study because of their various licensing processes and recent actions in the new NPP construction. Certain aspects of the aviation industry licensing process have also been studied and selected practices have been investigated as possibly suitable for use in nuclear licensing. Suitable features for SMR licensing are emphasized and suggested. The licensing features of the spent nuclear fuel deep repository along with similar features of SMR licensing are discussed. Since there are similar types of challenges of lengthy licensing time frames, as well as modular features to be taken into account in licensing, these two different nuclear industry fields can be compared. The main SMR features to take into account in licensing are: • Standardization of the design. • Modularity. • Mass production. • Serial construction. Modularity can be divided into two different categories: the first category is simply a single power plant unit constructed of independently engineered modules (e.g. construction process for Westinghouse AP-1000 NPP) and the second one a power plant composed of many reactor modules, which are manufactured in factories and installed as needed (e.g. NuScale Power SMR design). The deep underground repository

  15. Study of high-speed interaction processes between fluoropolymer projectiles and aluminum-based targets

    Institute of Scientific and Technical Information of China (English)

    Evgeny A. KHMELNIKOV; Alexey V. STYROV; Konstantin V. SMAGIN; Natalia S. KRAVCHENKO; Valery L. RUDENKO; Vladimir I. FALALEEV; Sergey S. SOKOLOV; Artem V. SVIDINSKY; Natalia F. SVIDINSKAYA

    2015-01-01

    The experimental results and numerical modeling of penetration process of fluoropolymer projectiles in aluminum-based targets are pre-sented. Analysis of mathematical models for interaction of elastoplastic projectile and target without taking additional energy released during interaction of fluoropolymer and aluminum into consideration is carried out. Energy fraction which is spent effectively on the increase in cavity volume is determined. The experimental and calculated results of penetration by combined and inert projectiles are compared.

  16. Transport safety record and measures taken for the United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program

    International Nuclear Information System (INIS)

    The United States Department of Energy (DOE), in consultation with the Department of State (DOS), adopted the Nuclear Weapons Non-proliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. As of January 2003, the Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) Acceptance Program has completed 25 shipments. Over 5,500 spent fuel assemblies from eligible research reactors throughout the world have been accepted into the United States under this program. Shipments are continuing on schedule, amidst a climate of heightened vigilance following the terrorist attacks of September 11, 2001. Recent guidance and an associated Order from the United States Nuclear Regulatory Commission (NRC) pertaining to security and safeguards issues deals directly with the transport of irradiated nuclear fuel. As the Acceptance Program consistently worked above and beyond basic regulatory requirements to provide additional safety enhancements in transporting spent nuclear fuel, the new guidance did not adversely effect the Program. Other global SNF shipping campaigns and the recent recommendation to proceed with the licensing application process for Yucca Mountain as the geologic repository for spent fuel have increased media and public interest in SNF transport. The Acceptance Policy is planned to expire in May 2006, per the 1996 Programmatic Record of Decision on the Policy, but some shipments will continue until May 2009. Currently, there are no plans to renew the Acceptance Policy. This paper examines the safety record of the Acceptance Program, and discusses the measures that have been taken to ensure future shipments are safe and uneventful. (author)

  17. Precisely determined the spent nuclear fuel antineutrino flux and spectrum for Daya Bay antineutrino experiment

    CERN Document Server

    Ma, X B; Chen, Y X; Zhong, W L; An, F P

    2015-01-01

    Spent nuclear fuel (SNF) antineutrino flux is an important source of uncertainties for a reactor neutrino flux prediction. However, if one want to determine the contribution of spent fuel, many data are needed, such as the amount of spent fuel in the pool, the time after discharged from the reactor core, the burnup of each assembly, and the antineutrino spectrum of the isotopes in the spend fuel. A method to calculate the contribution of SNF is proposed in this study. In this method, reactor simulation code verified by experiment have been used to simulate the fuel depletion by taking into account more than 2000 isotopes and fission products, the quantity of SNF in each six spend fuel pool, and the antineutrino spectrum of SNF varying with time after SNF discharged from core. Results show that the contribution of SNF to the total antineutrino flux is about 0.26%~0.34%, and the shutdown impact is about 20%. The SNF spectrum would distort the softer part of antineutrino spectra, and the maximum contribution fro...

  18. Experimental observations on electrorefining spent nuclear fuel in molten LiCl-KCl/liquid cadmium system.

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, T. A.; Laug, D. V.; Li, S. X.; Sofu, T.

    1999-07-14

    Argonne National Laboratory (ANL) is currently performing a demonstration program for the Department of Energy (DOE) which processes spent nuclear fuel from the Experimental Breeder Reactor (EBR-II). One of the key steps in this demonstration program is electrorefining of the spent fuel in a molten LiCl-KCl/liquid cadmium system using a pilot scale electrorefiner (Mk-IV ER). This article summarizes experimental observations and engineering aspects for electrorefining spent fuel in the molten LiCl-KCl/liquid cadmium system. It was found that the liquid cadmium pool acted as an intermediate electrode during the electrorefining process in the ER. The cadmium level was gradually decreased due to its high vapor pressure and vaporization rate at the ER operational temperature. The low cadmium level caused the anode assembly momentarily to touch the ER vessel hardware, which generated a periodic current change at the salt/cathode interface and improved uranium recovery efficiency for the process. The primary current distributions calculated by numerical simulations were used in interpreting the experimental results.

  19. Measurement of plutonium in spent nuclear fuel by self-induced x-ray fluorescence

    Energy Technology Data Exchange (ETDEWEB)

    Hoover, Andrew S [Los Alamos National Laboratory; Rudy, Cliff R [Los Alamos National Laboratory; Tobin, Steve J [Los Alamos National Laboratory; Charlton, William S [Los Alamos National Laboratory; Stafford, A [TEXAS A& M; Strohmeyer, D [TEXAS A& M; Saavadra, S [ORNL

    2009-01-01

    Direct measurement of the plutonium content in spent nuclear fuel is a challenging problem in non-destructive assay. The very high gamma-ray flux from fission product isotopes overwhelms the weaker gamma-ray emissions from plutonium and uranium, making passive gamma-ray measurements impossible. However, the intense fission product radiation is effective at exciting plutonium and uranium atoms, resulting in subsequent fluorescence X-ray emission. K-shell X-rays in the 100 keV energy range can escape the fuel and cladding, providing a direct signal from uranium and plutonium that can be measured with a standard germanium detector. The measured plutonium to uranium elemental ratio can be used to compute the plutonium content of the fuel. The technique can potentially provide a passive, non-destructive assay tool for determining plutonium content in spent fuel. In this paper, we discuss recent non-destructive measurements of plutonium X-ray fluorescence (XRF) signatures from pressurized water reactor spent fuel rods. We also discuss how emerging new technologies, like very high energy resolution microcalorimeter detectors, might be applied to XRF measurements.

  20. Storing the Spent Nuclear Fuel in Dry Casks Licensed for a Century as an Alternative to Recycling Solution

    Science.gov (United States)

    Milincic, Radovan

    2010-02-01

    Management of spent nuclear power reactor fuels is one of the most urgent problems in nuclear technology. Yearly production of new spent fuel is in the range of thousands of tons, topping a couple of hundred thousand tons of spent fuel already. This material is extremely radioactive and currently there is no adequate international policy, control or management regarding it. I propose here an intermediate term solution to this problem, which will be technologically and economically sustainable: interim spent-fuel storage as an alternative to reprocessing. The reprocessing inherently increases the net amount of the plutonium, which can be used for production of nuclear arms. Moreover, it is an expensive process with the net effect of producing different type of radioactive waste. In particular, the development of a dry cask for nuclear waste storage on site and transport, licensed for a period of hundred years would provide a significantly less expensive solution in the recent future, giving a needed relief to crowded spent-fuel storage pools. Currently in the U.S, NRC licenses existing storage casks for 20 years; and licenses for some of the dry cask storage facilities in the U.S. are about to expire. The extended life dry casks will provide sufficient intermediate period toward a more efficient and/or technologically advanced solution for spent fuel. )

  1. Long-term management of Canada's spent nuclear fuel: the nuclear waste management organizations recommendation to government

    International Nuclear Information System (INIS)

    Full text: Like many countries with nuclear power programs, Canada is in the process of addressing the long-term management of its spent fuel. The Nuclear Waste Management Organization (NWMO) was tasked through federal legislation to conduct a three-year study of approaches for the long-term management of spent fuel, and to recommend a preferred approach to the Government of Canada. Legislation required NWMO to compare at least three approaches -approaches based on deep geological disposal in the Canadian Shield, storage at nuclear reactor sites, and centralized storage either above or below ground. In assessing the options, NWMO sought a recommendation that would be socially acceptable, technically sound, environmentally responsible and economically feasible. The study drew on a vast base of social, technical, engineering, and financial research, and included an extensive engagement program with the public and Aboriginal peoples. The recommendation emerged from a collaborative dialogue with specialists and citizens, for an approach that is built on sound science and technology and responsive to citizen values. NWMO submitted its completed options study, with recommendation, to the Government in November 2005. NWMO has proposed an alternative approach, Adaptive Phased Management, which has as its key attributes: central containment and isolation of spent fuel in a deep repository, in an appropriate geological formation; contingency provision for central shallow storage; monitoring and retrievability; and a staged, adaptive process of concept implementation, reflecting the complex nature of the task and the desire of citizens to proceed through cautious, deliberate steps of technical demonstration and social acceptance. This paper will review: 1) the development of the assessment framework for comparing the technical options, which incorporated social and ethical considerations expressed by citizens; 2) findings of the assessment; and 3) features of the proposed

  2. Site selection - location of the repository for spent nuclear fuel; Platsval - lokalisering av slutfoervaret foer anvaent kaernbraensle

    Energy Technology Data Exchange (ETDEWEB)

    2010-12-15

    This document describes the localization work and SKB's choice of site for the repository. Furthermore, SKB's basis and rationale for the decisions taken during the work are reported. The document is Appendix PV of applications under the Nuclear Activities Act and the Environmental Code to both build and operate an encapsulation plant adjacent to the central interim storage facility for spent nuclear fuel in Oskarshamn, and to construct and operate a disposal facility for spent nuclear fuel at Forsmark in Oesthammar municipality

  3. Burnup determination and age dating of spent nuclear fuel using noble gas isotopic analysis

    International Nuclear Information System (INIS)

    During the chopping and dissolving phases of reprocessing, gases (such as tritium, krypton, xenon, iodine, carbon dioxide, nitrogen oxide, and steam) are released. These gases are traditionally transferred to a gas-treatment system for treatment, release, and/or recycle. Because of their chemically inert nature, the xenon and krypton noble gases are generally released directly into the loser atmosphere through the facility's stack. These gases (being fission products) contain information about the fuel being reprocessed and may prove a valuable monitor of reprocessing activities. Two properties of the fuel that may prove valuable from a safeguards standpoint are the fuel burnup and the fuel age (or time since discharge from the reactor). Both can be used to aid in confirming declared activities, and the burnup is generally indicative of the usability of the fuel for fabricating nuclear explosives. A study has been ongoing at Los Alamos National Laboratory to develop a methodology to determine spent-fuel parameters from measured xenon and/or krypton isotopic ratios on-stack at reprocessing facilities. This study has resulted in the generation of the NOVA data analysis code, which links to a comprehensive database of reactor physics parameters (calculated using the Monteburns 3.01 code system). NOVA has been satisfactorily tested for burnup determination of weapons-grade fuel from a US production reactor. Less effort has been spent quantifying NOVA's ability to predict burnup and fuel age for power reactor fuel. The authors describe the results predicted by NOVA for xenon and krypton isotopic ratios measured after the dissolution of spent-fuel samples from the Borssele reactor. The Borssele reactor is a 450-MW(electric) pressurized water reactor (PWR) consisting of 15 x 15 KWU assemblies. The spent-fuel samples analyzed were single fuel rods removed from one assembly and dissolved at the La Hague reprocessing facility. The assembly average burnup was estimated at 32

  4. Experimental Investigation of Burnup Credit for Safe Transport, Storage, and Disposal of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Harms, Gary A.; Helmick, Paul H.; Ford, John T.; Walker, Sharon A.; Berry, Donald T.; Pickard, Paul S.

    2004-04-01

    This report describes criticality benchmark experiments containing rhodium that were conducted as part of a Department of Energy Nuclear Energy Research Initiative project. Rhodium is an important fission product absorber. A capability to perform critical experiments with low-enriched uranium fuel was established as part of the project. Ten critical experiments, some containing rhodium and others without, were conducted. The experiments were performed in such a way that the effects of the rhodium could be accurately isolated. The use of the experimental results to test neutronics codes is demonstrated by example for two Monte Carlo codes. These comparisons indicate that the codes predict the behavior of the rhodium in the critical systems within the experimental uncertainties. The results from this project, coupled with the results of follow-on experiments that investigate other fission products, can be used to quantify and reduce the conservatism of spent nuclear fuel safety analyses while still providing the necessary level of safety.

  5. Experimental investigation of burnup credit for safe transport, storage, and disposal of spent nuclear fuel.

    Energy Technology Data Exchange (ETDEWEB)

    Berry, Donald T.; Harms, Gary A.; Ford, John T.; Walker, Sharon Ann; Helmick, Paul H.; Pickard, Paul S.

    2004-04-01

    This report describes criticality benchmark experiments containing rhodium that were conducted as part of a Department of Energy Nuclear Energy Research Initiative project. Rhodium is an important fission product absorber. A capability to perform critical experiments with low-enriched uranium fuel was established as part of the project. Ten critical experiments, some containing rhodium and others without, were conducted. The experiments were performed in such a way that the effects of the rhodium could be accurately isolated. The use of the experimental results to test neutronics codes is demonstrated by example for two Monte Carlo codes. These comparisons indicate that the codes predict the behavior of the rhodium in the critical systems within the experimental uncertainties. The results from this project, coupled with the results of follow-on experiments that investigate other fission products, can be used to quantify and reduce the conservatism of spent nuclear fuel safety analyses while still providing the necessary level of safety.

  6. Enrico Fermi Fast Reactor Spent Nuclear Fuel Criticality Calculations: Degraded Mode

    International Nuclear Information System (INIS)

    The objective of this calculation is to characterize the nuclear criticality safety concerns associated with the codisposal of the Department of Energy's (DOE) Enrico Fermi (EF) Spent Nuclear Fuel (SNF) in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP) and placed in a Monitored Geologic Repository (MGR). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (keff) for the degraded mode internal configurations of the codisposal WP. The results of this calculation and those of Ref. 8 will be used to evaluate criticality issues and support the analysis that will be performed to demonstrate the viability of the codisposal concept for the Monitored Geologic Repository

  7. A study of antineutrino spectra from spent nuclear fuel at Daya Bay

    Institute of Scientific and Technical Information of China (English)

    ZHOU Bin; RUAN Xi-Chao; NIE Yang-Bo; ZHOU Zu-Ying; AN Feng-Peng; CAO Jun

    2012-01-01

    The Daya Bay Reactor Antineutrino Experiment is designed to determine the as yet unknown neutrino mixing angle,θ13,by measuring the disappearance of electron antineutrinos from several nuclear reactor cores.The projected sensitivity in sin2(2θ13) of better than 0.01 at a 90% CL should be achieved after three years of data-taking.Antineutrinos emitted from spent nuclear fuel (SNF) distort the soft part of the energy spectrum.In this article,a calculation of the antineutrino spectra from the long-life isotopes in SNF is performed.A non-equilibrium generation of long half-life isotopes during the running time of the reactor is also analyzed.Finally,we show that the antineutrino event rate contribution from SNF,which has been stored in the SNF pool for several years,may be non-negligible.

  8. Interim Storage of Spent Nuclear Fuel before Final Disposal in Germany - Regulator's view

    International Nuclear Information System (INIS)

    For spent nuclear fuel management in Germany the concept of dry interim storage in dual purpose casks before direct disposal is applied. The Federal Office for Radiation Protection (BfS) is the competent authority for licensing of interim storage facilities. The competent authority for surveillance of operation is the responsible authority of the respective federal state (Land). Currently operation licenses for storage facilities have been granted for a storage time of 40 years and are based on safety demonstrations for all safety issues as safe enclosure, shielding, sub-criticality and decay heat removal under consideration of operation conditions. In addition, transportability of the casks for the whole storage period has to be provided. Due to current delay in site selection and exploration of a disposal site, an extension of the storage time beyond 40 years could be needed. This will cause appropriate actions by the licensee and the competent authorities as well. A brief description of the regulatory base of licensing and surveillance of interim storage is given from the regulators view. Furthermore the current planning for final disposal of spent nuclear fuel and high level waste and its interconnections between storage and disposal concepts are shortly explained. Finally the relevant aspects for licensing of extended storage time beyond 40 years will be discussed. Current activities on this issue, which have been initiated by the Federal Government, will be addressed. On the regulatory side a review and amendment of the safety guideline for interim storage of spent fuel has been performed and the procedure of periodic safety review is being implemented. A guideline for implementing an ageing management programme is available in a draft version. Regarding safety of long term storage a study focussing on the identification and evaluation of long term effects as well as gaps of knowledge has been finished in 2010. A continuation and update is currently underway

  9. Simulation of differential die-away instrument's response to asymmetrically burned spent nuclear fuel

    Science.gov (United States)

    Martinik, Tomas; Henzl, Vladimir; Grape, Sophie; Svärd, Staffan Jacobsson; Jansson, Peter; Swinhoe, Martyn T.; Tobin, Stephen J.

    2015-07-01

    Previous simulation studies of Differential Die-Away (DDA) instrument's response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument's response to interrogation of asymmetrically burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs. The results of this study suggest that DDA instrument response depends on the position of the individual neutron detectors and in fact can be split in two modes. The first mode, measured by the back detectors, is not significantly sensitive to the spatial distribution of fissile isotopes and neutron absorbers, but rather reflects the total amount of both contributors as in the cases of symmetrically burned SFAs. In contrary, the second mode, measured by the front detectors, yields certain sensitivity to the orientation of the asymmetrically burned SFA inside the assaying instrument. This study thus provides evidence that the DDA instrument can potentially be utilized as necessary in both ways, i.e. a quick determination of the average SFA characteristics in a single assay, as well as a more detailed characterization involving several DDA observables through assay of the SFA from all of its four sides that can possibly map the burn-up distribution and/or identify diversion or

  10. Simulation of differential die-away instrument’s response to asymmetrically burned spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Martinik, Tomas, E-mail: tomas.martinik@physics.uu.se [Department of Physics and Astronomy, Uppsala University, Box 516 Sweden, SE-75120 Uppsala (Sweden); Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM, 87545 (United States); Henzl, Vladimir [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM, 87545 (United States); Grape, Sophie; Svärd, Staffan Jacobsson; Jansson, Peter [Department of Physics and Astronomy, Uppsala University, Box 516 Sweden, SE-75120 Uppsala (Sweden); Swinhoe, Martyn T. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM, 87545 (United States); Tobin, Stephen J. [Department of Physics and Astronomy, Uppsala University, Box 516 Sweden, SE-75120 Uppsala (Sweden); Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM, 87545 (United States); Swedish Nuclear Fuel and Waste Management Company, Blekholmstorget 30, Box 250, SE-101 24 Stockholm (Sweden)

    2015-07-11

    Previous simulation studies of Differential Die‐Away (DDA) instrument’s response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument’s response to interrogation of asymmetrically burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs. The results of this study suggest that DDA instrument response depends on the position of the individual neutron detectors and in fact can be split in two modes. The first mode, measured by the back detectors, is not significantly sensitive to the spatial distribution of fissile isotopes and neutron absorbers, but rather reflects the total amount of both contributors as in the cases of symmetrically burned SFAs. In contrary, the second mode, measured by the front detectors, yields certain sensitivity to the orientation of the asymmetrically burned SFA inside the assaying instrument. This study thus provides evidence that the DDA instrument can potentially be utilized as necessary in both ways, i.e. a quick determination of the average SFA characteristics in a single assay, as well as a more detailed characterization involving several DDA observables through assay of the SFA from all of its four sides that can possibly map the burn-up distribution and/or identify diversion or

  11. Fusion solution to dispose of spent nuclear fuel, transuranic elements, and highly enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Yousry E-mail: gohar@anl.gov

    2001-11-01

    The disposal of the nuclear spent fuel, the transuranic elements, and the highly enriched uranium represents a major problem under investigation by the international scientific community to identify the most promising solutions. The investigation of this paper focused on achieving the top rated solution for the problem, the elimination goal, which requires complete elimination for the transuranic elements or the highly enriched uranium, and the long-lived fission products. To achieve this goal, fusion blankets with liquid carrier, molten salts or liquid metal eutectics, for the transuranic elements and the uranium isotopes are utilized. The generated energy from the fusion blankets is used to provide revenue for the system. The long-lived fission products are fabricated into fission product targets for transmutation utilizing the neutron leakage from the fusion blankets. This paper investigated the fusion blanket designs for small fusion devices and the system requirements for such application. The results show that 334 MW of fusion power from D-T plasma for 30 years with an availability factor of 0.75 can dispose of the 70,000 tons of the U.S. inventory of spent nuclear fuel generated up to the year 2015. In addition, this fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future.

  12. Oxidative dissolution of spent nuclear fuel in aqueous alkaline solutions - An alternative to the Purex process?

    Energy Technology Data Exchange (ETDEWEB)

    Runde, Wolfgang; Peper, Shane; Brodnax, Lia; Crooks, William; Zehnder, Ralph; Jarvinen, Gordon

    2004-07-01

    As an alternative to acidic reprocessing of spent nuclear, oxidative dissolution of UO{sub 2} into aqueous alkaline solutions and subsequent separation of fission products is considered. The efficacy of such a method is limited by the kinetics of the UO{sub 2} dissolution and the capacity of alkaline solutions for dissolved U(VI) species. We performed a series of dissolution studies on UO{sub 2} and U{sub 3}O{sub 8} in aqueous alkaline solutions applying various oxidants. Among the oxidative agents commonly used to transform low-valence actinides into their higher oxidation states, H{sub 2}O{sub 2} has proven to be the most effective in basic media. Consequently, we investigated the dissolution of UO{sub 2} and U{sub 3}O{sub 8} in NaOH-H{sub 2}O{sub 2} and Na{sub 2}CO{sub 3}-H{sub 2}O{sub 2} solutions and determined the dissolution kinetics as a function of peroxide and hydroxide (carbonate) concentrations. Methods to remove fission products, e.g., Cs, Sr, Ba and Zr, from alkaline solutions will be evaluated based upon their decontamination factors. We will discuss the feasibility of using chemically oxidizing alkaline solutions as an alternative spent nuclear fuel reprocessing method based on results from experimental quantitative investigations. (authors)

  13. Impact Analyses and Tests of Concrete Overpacks of Spent Nuclear Fuel Storage Casks

    International Nuclear Information System (INIS)

    A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches, those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters

  14. Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

    International Nuclear Information System (INIS)

    Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training

  15. Sampling and analysis plan for the preoperational environmental survey of the spent nuclear fuel project facilities

    International Nuclear Information System (INIS)

    This sampling and analysis plan will support the preoperational environmental monitoring for construction, development, and operation of the Spent Nuclear Fuel (SNF) Project facilities, which have been designed for the conditioning and storage of spent nuclear fuels; particularly the fuel elements associated with the operation of N-Reactor. The SNF consists principally of irradiated metallic uranium, and therefore includes plutonium and mixed fission products. The primary effort will consist of removing the SNF from the storage basins in K East and K West Areas, placing in multicanister overpacks, vacuum drying, conditioning, and subsequent dry vault storage in the 200 East Area. The primary purpose and need for this action is to reduce the risks to public health and safety and to the environment. Specifically these include prevention of the release of radioactive materials into the air or to the soil surrounding the K Basins, prevention of the potential migration of radionuclides through the soil column to the nearby Columbia River, reduction of occupational radiation exposure, and elimination of the risks to the public and to workers from the deterioration of SNF in the K Basins

  16. Strategic Minimization of High Level Waste from Pyroprocessing of Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    The pyroprocessing of spent nuclear fuel results in two high-level waste streams--ceramic and metal waste. Ceramic waste contains active metal fission product-loaded salt from the electrorefining, while the metal waste contains cladding hulls and undissolved noble metals. While pyroprocessing was successfully demonstrated for treatment of spent fuel from Experimental Breeder Reactor-II in 1999, it was done so without a specific objective to minimize high-level waste generation. The ceramic waste process uses 'throw-away' technology that is not optimized with respect to volume of waste generated. In looking past treatment of EBR-II fuel, it is critical to minimize waste generation for technology developed under the Global Nuclear Energy Partnership (GNEP). While the metal waste cannot be readily reduced, there are viable routes towards minimizing the ceramic waste. Fission products that generate high amounts of heat, such as Cs and Sr, can be separated from other active metal fission products and placed into short-term, shallow disposal. The remaining active metal fission products can be concentrated into the ceramic waste form using an ion exchange process. It has been estimated that ion exchange can reduce ceramic high-level waste quantities by as much as a factor of 3 relative to throw-away technology

  17. Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

    Energy Technology Data Exchange (ETDEWEB)

    D.R. Jackson; G.R. Kiebel

    1999-08-24

    Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

  18. Consortium NCS/GNS: Disposal of spent nuclear fuel from the DKFZ (Heidelberg, Germany)

    Energy Technology Data Exchange (ETDEWEB)

    Vallentin, R. [GNS Gesellschaft fuer Nuklear-Service mbH, Hollestr. 7 A, D- 45127 Essen, WTI Wissenschaftlich- Technische Ingenieurberatung GmbH, K.-H.-Beckurtsstr. 8, 52428 Juelich (Germany); Juenger, B. [DKFZ Deutsches Krebsforschungszentrum, Im Neuenheimer Feld 280, D-69120 Heidelberg (Germany); Schmidt, T. [NCS Nuclear Cargo and Service, Rodenbacher Chaussee 6, D-63457 Hanau (Germany)

    2002-07-01

    Before the decommissioning of the TRIGA HD II reactor of the 'Deutsches Krebsforschungszentrum' in Heidelberg (DKFZ) the spent nuclear fuel (126 TRIGA- F/A) had to be discharged from the reactor and shipped in 2001 to the United States. The Consortium NCS/GNS together with the DKFZ carried out this task including the following essential details: Implementation of structural measures inside the reactor building. Development and fabrication of loading units for the FIA. Development and fabrication of a transfer system for the loading units. Obtaining new German and US certifications for the casks. Loading of the casks. Shipment of the casks from DKFZ to the United States by road and sea. The successful completion of this task once again proved the operational experience of the Consortium NCS/GNS which will be useful for further worldwide shipments of spent nuclear fuel from research reactors. e.g. shipments from Europe, South America and Australia to the United States and France. (author)

  19. The Chemistry of Spent Nuclear Fuel From X-Ray Absorption Spectroscopy

    International Nuclear Information System (INIS)

    Present and future nuclear fuel cycles will require an understanding of the complex chemistry of trace fission products and transuranium actinides in spent nuclear fuel (SNF). Because of the unique analytical challenges presented by SNF to the materials scientist, many of its fundamental physical and chemical properties remain poorly understood, especially on the microscopic scale. Such an understanding of the chemical states of radionuclides in SNF would benefit development of technologies for fuel monitoring, fuel performance improvement and modeling, fuel reprocessing, and spent fuel storage and disposal. We have recently demonstrated the use of synchrotron x-ray absorption spectroscopy (XAS) to examine crystal chemical properties of actinides and fission products in extracted specimens of SNF. Information obtained includes oxidation state, chemical bond coordination, and quantitative elemental concentration and distribution. We have also used XAS in a scanning mode to obtain x-ray spectral micrographs with resolution approaching 1 micron. A brief overview of the technique will be presented, along with findings on uranium, plutonium, neptunium, technetium, and molybdenum in commercial PWR SNF specimens

  20. Sampling and analysis plan for the preoperational environmental survey of the spent nuclear fuel project facilities

    Energy Technology Data Exchange (ETDEWEB)

    MITCHELL, R.M.

    1999-04-01

    This sampling and analysis plan will support the preoperational environmental monitoring for construction, development, and operation of the Spent Nuclear Fuel (SNF) Project facilities, which have been designed for the conditioning and storage of spent nuclear fuels; particularly the fuel elements associated with the operation of N-Reactor. The SNF consists principally of irradiated metallic uranium, and therefore includes plutonium and mixed fission products. The primary effort will consist of removing the SNF from the storage basins in K East and K West Areas, placing in multicanister overpacks, vacuum drying, conditioning, and subsequent dry vault storage in the 200 East Area. The primary purpose and need for this action is to reduce the risks to public health and safety and to the environment. Specifically these include prevention of the release of radioactive materials into the air or to the soil surrounding the K Basins, prevention of the potential migration of radionuclides through the soil column to the nearby Columbia River, reduction of occupational radiation exposure, and elimination of the risks to the public and to workers from the deterioration of SNF in the K Basins.

  1. SNG experience in dry storage of spent fuel and nuclear waste

    International Nuclear Information System (INIS)

    SGN has gained considerable experience in the design and construction of interim storage facilities for spent fuel and various nuclear waste, and can therefore propose single product and multi-product facilities capable of accommodating all types of waste in a single structure. The pooling of certain functions (transport cask reception, radiation protection) and the choice of optimized technologies to answer the specific needs of the Clients (automatic transfer by shielded cask or nuclearized crane), the use of the same type of well to cool the heat releasing packages (glass canisters, fuel elements) make it possible to propose industrially proven and cost-effective solutions. The studies conducted by SGN, on behalf of the Dutch company COVRA (Centrale Organisatie Voor Radioactif Afval), offer an example of the application of this new concept. This paper first presents the SGN experience through a short description of reference storage facilities for various types of products (MLW, HLW and Spent Fuel). It goes on with a typical application to show how these proven technologies are combined to obtain single product or multi-product facilities tailored to the Client's specific requirements. (authors)

  2. SGN experience in dry storage of spent fuel and nuclear waste

    International Nuclear Information System (INIS)

    SGN has gained considerable experience in the design and construction of interim storage facilities for spent fuel and various nuclear waste, and can therefore propose single product and multiproduct facilities capable of accommodating all types of waste in a single structure. The pooling of certain functions (transport cask reception, radiation protection) and the choice of optimized technologies to answer the specific needs of the Clients (automatic transfer by shielded cask or nuclearized crane), the use of the same type of well to cool the heat releasing packages (glass canisters, fuel elements) make it possible to propose industrially proven and cost-effective solutions. The studies conducted by SGN, on behalf of the Dutch company COVRA (Centrale Organisatie Voor Radioactif Afval), offer an example of the application of this new concept. This paper first presents the SGN experience through a short description of reference storage facilities for various types of products (MLW, HLW and Spent Fuel). It goes on with a typical application to show how these proven technologies are combined to obtain single product or multiproduct facilities tailored to the Client's specific requirements

  3. Effects of temperature on concrete cask in a dry storage facility for spent nuclear fuels

    International Nuclear Information System (INIS)

    In the dry storage of spent nuclear fuels,concrete cask serves both as a shielding and a structural containment. The concrete in the storage facility is expected to endure the decay heat of the spent nuclear fuel during its service life. Thus, effects of the sustaining high temperature on concrete material need be evaluated for safety of the dry storage facility. In this paper, we report an experimental program aimed at investigating possible high temperature effects on properties of concrete, with emphasis on the mechanical stability, porosity,and crack-resisting ability of concrete mixes prepared using various amounts of Portland cement, fly ash, and blast furnace slag. The experimental results obtained from concrete specimens exposed to a temperature of 94 degree C for 90 days indicate that: (1) compressive strength of the concrete remains practically unchanged; (2) the ultrasonic pulse velocity, and dynamic modulus of elasticity of the concrete decrease in early stage of the high-temperature exposure,and gradually become stable with continuing exposure; (3) shrinkage of concrete mixes exhibits an increase in early stage of the exposure and does not decrease further with time; (4) concrete mixes containing pozzolanic materials,including fly ash and blast furnace slag, show better temperature-resisting characteristics than those using only Portland cement. (authors)

  4. U.S. spent nuclear fuel management: Political, fiscal, and technical feasibility

    International Nuclear Information System (INIS)

    Successful U.S. spent nuclear fuel management policy must satisfy political, fiscal, and technical constraints. Technical requirements have been thoroughly investigated in the United States and Nordic countries for volcanic tuff, salt, and granite. Fiscal planning requires an inflation-adjusted revenue stream and predictable real interest rate earnings on fund balances. A prompt solution satisfying political constraints requires compromise between the overlapping but distinct goals of seven different sets of interests at the federal level. Absent such compromise, there will be delay until sufficient support for one of three strategies evolves: (1) force the Yucca Mountain repository on Nevada, (2) open a centralized storage facility without coupling to repository licensing, or (3) follow a “consent-based” process for repository licensing. Formulations of each of these strategies to overcome impediments to their success are described. - Highlights: • U.S. spent nuclear fuel policy has been at an impasse. • No compromise on Yucca Mountain is at hand. • Yucca only, new repository, or no repository are options. • Success with each is more likely with sounder financing

  5. Centrifugal microfluidic platform for radiochemistry: potentialities for the chemical analysis of nuclear spent fuels.

    Science.gov (United States)

    Bruchet, Anthony; Taniga, Vélan; Descroix, Stéphanie; Malaquin, Laurent; Goutelard, Florence; Mariet, Clarisse

    2013-11-15

    The use of a centrifugal microfluidic platform is for the first time reported as an alternative to classical chromatographic procedures for radiochemistry. The original design of the microfluidic platform has been thought to fasten and simplify the prototyping process with the use of a circular platform integrating four rectangular microchips made of thermoplastic. The microchips, dedicated to anion-exchange chromatographic separations, integrate a localized monolithic stationary phase as well as injection and collection reservoirs. The results presented here were obtained with a simplified simulated nuclear spent fuel sample composed of non-radioactive isotopes of Europium and Uranium, in proportion usually found for uranium oxide nuclear spent fuel. While keeping the analytical results consistent with the conventional procedure (extraction yield for Europium of ≈97%), the use of the centrifugal microfluidic platform allowed to reduce the volume of liquid needed by a factor of ≈250. Thanks to their unique "easy-to-use" features, centrifugal microfluidic platforms are potential successful candidates for the downscaling of chromatographic separation of radioactive samples (automation, multiplexing, easy integration in glove-boxes environment and low cost of maintenance).

  6. Analysis of dose consequences arising from the release of spent nuclear fuel from dry storage casks.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G.; Morrow, Charles.

    2013-01-01

    The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.

  7. Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL; Mueller, Don [ORNL; Goluoglu, Sedat [ORNL; Hollenbach, Daniel F [ORNL; Fox, Patricia B [ORNL

    2007-10-01

    The purpose of this calculation report, Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel, is to validate the computational method used to perform postclosure criticality calculations. The validation process applies the criticality analysis methodology approach documented in Section 3.5 of the Disposal Criticality Analysis Methodology Topical Report. The application systems for this validation consist of waste packages containing transport, aging, and disposal canisters (TAD) loaded with commercial spent nuclear fuel (CSNF) of varying assembly types, initial enrichments, and burnup values that are expected from the waste stream and of varying degree of internal component degradation that may occur over the 10,000-year regulatory time period. The criticality computational tool being evaluated is the general-purpose Monte Carlo N-Particle (MCNP) transport code. The nuclear cross-section data distributed with MCNP 5.1.40 and used to model the various physical processes are based primarily on the Evaluated Nuclear Data File/B Version VI (ENDF/B-VI) library. Criticality calculation bias and bias uncertainty and lower bound tolerance limit (LBTL) functions for CSNF waste packages are determined based on the guidance in ANSI/ANS 8.1-1998 (Ref. 4) and ANSI/ANS 8.17-2004 (Ref. 5), as described in Section 3.5.3 of Ref. 1. The development of this report is consistent with Test Plan for: Range of Applicability and Bias Determination for Postclosure Criticality. This calculation report has been developed in support of licensing activities for the proposed repository at Yucca Mountain, Nevada, and the results of the calculation may be used in the criticality evaluation for CSNF waste packages based on a conceptual TAD canister.

  8. Study of a brazilian cask and its installation for PWR spent nuclear fuel dry storage

    International Nuclear Information System (INIS)

    Spent nuclear fuel (SNF) is removed from the nuclear reactor after the depletion on efficiency in generating energy. After the withdrawal from the reactor core, the SNF is temporarily stored in pools at the same site of the reactor. At this time, the generated heat and the short and medium lived radioactive elements decay to levels that allow removing SNF from the pool and sending it to temporary dry storage. In that phase, the fuel needs to be safely and efficiently stored, and then, it can be retrieved in a future, or can be disposed as radioactive waste. The amount of spent fuel increases annually and, in the next years, will still increase more, because of the construction of new nuclear plants. Today, the number of new facilities back up to levels of the 1970's, since it is greater than the amount of decommissioning in old installations. As no final decision on the back-end of the nuclear fuel cycle is foreseen in the near future in Brazil, either to recover the SNF or to consider it as radioactive waste, this material has to be isolated in some type of storage model existing around the world. In the present study it is shown that dry SNF storage is the best option. A national cask model for SNF as well these casks storage installation are proposed. It is a multidisciplinary study in which the engineering conceptual task was developed and may be applied to national SNF removed from the Brazilian power reactors, to be safely stored for a long time until the Brazilian authorities will decide about the site for final disposal. (author)

  9. Analysis of civil liability for nuclear damage with emphasis on the insurance of liability for the transport of fresh nuclear fuel and spent nuclear fuel

    International Nuclear Information System (INIS)

    The issue of civil liability for nuclear damage and the mandatory coverage of this liability by insurance or by the other financial means is a topic for experts discussions, particularly due to the existing differences in opinion. The article outlines the major provisions of the Vienna Convention on Civil Liability for Nuclear Damage (1963) and presents a list of parties to the Convention from among Central and Eastern European countries. In the second part, the paper deals with the various options during the transportation of spent and fresh nuclear fuel from the viewpoint of the regulatory body and with regard to applicable domestic and international legislation. (author)

  10. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact Statement, Volume 1, Appendix C

    OpenAIRE

    United States Department of Energy

    1995-01-01

    The U.S. Department of Energy (DOE) is engaged in two related decisionmaking processes concerning: (1) the transportation, receipt, processing, and storage of spent nuclear fuel (SNF) at the DOE Idaho National Engineering Laboratory (INEL) which will focus on the next 10 years; and (2) programmatic decisions on future spent nuclear fuel management which will emphasize the next 40 years. DOE is analyzing the environmental consequences of these spent nuclear fuel management actions in this tw...

  11. A nuclear fuel cycle system dynamic model for spent fuel storage options

    International Nuclear Information System (INIS)

    Highlights: • Used nuclear fuel management requires a dynamic system analysis study due to its socio-technical complexity. • Economic comparison of local, regional, and national storage options is limited due to the public financial information. • Local and regional options of used nuclear fuel management are found to be the most economic means of storage. - Abstract: The options for used nuclear fuel storage location and affected parameters such as economic liabilities are currently a focus of several high level studies. A variety of nuclear fuel cycle system analysis models are available for such a task. The application of nuclear fuel cycle system dynamics models for waste management options is important to life-cycle impact assessment. The recommendations of the Blue Ribbon Committee on America’s Nuclear Future led to increased focus on long periods of spent fuel storage [1]. This motivated further investigation of the location dependency of used nuclear fuel in the parameters of economics, environmental impact, and proliferation risk. Through a review of available literature and interactions with each of the programs available, comparisons of post-reactor fuel storage and handling options will be evaluated based on the aforementioned parameters and a consensus of preferred system metrics and boundary conditions will be provided. Specifically, three options of local, regional, and national storage were studied. The preliminary product of this research is the creation of a system dynamics tool known as the Waste Management Module (WMM) which provides an easy to use interface for education on fuel cycle waste management economic impacts. Initial results of baseline cases point to positive benefits of regional storage locations with local regional storage options continuing to offer the lowest cost

  12. Final disposal of spent nuclear fuel - regulatory system and roles of different actors during the decision process

    Energy Technology Data Exchange (ETDEWEB)

    2009-03-15

    In November 2006 Swedish Nuclear Fuels Co. applied for a license to build a plant for encapsulation of spent nuclear fuels at Oskarshamn, Sweden. The company also have plans to apply, in 2009, for a license to construct a underground repository for spent nuclear fuels. KASAM arranged a seminar in November 2006 in order to describe and discuss the licensing rules and regulations and the roles of different parties in the decision making. Another objective of the seminar was to point out possible ambiguities in this process. Another interesting question under discussion was in what ways the basic data for the decision should be produced. The seminar covered the part of the process beginning with the application for a license and ending with the government approval/rejection of the application. Most time was spent on the legal aspects of the process

  13. Summary of the spent nuclear fuel transportation workshop for state officials, October 22, 1985, Springfield, Illinois. Final report

    International Nuclear Information System (INIS)

    On October 22, 1985, thirty-two state officials representing eighteen states attended a workshop on the transportation of spent nuclear fuel hosted by the Illinois Department of Nuclear Safety and co-sponsored by the US Department of Energy. The letter of invitation is attached to this report along with the agenda for the workshop. Also attached is the attendance list. The purposes of the workshop were to explain Illinois' inspection, escort and emergency preparedness program for spent nuclear fuel to officials from other states and to discuss the needs of the future national system when DOE begins shipping spent fuel to either a repository or a monitored retrievable storage facility. Summaries of the presentations and the reports from the discussion groups are presented

  14. The impact of a final disposal facility for spent nuclear fuel on a municipality's image

    International Nuclear Information System (INIS)

    The study comprised on one hand a nationwide telephone interview (totally 800 interviews) aimed at mapping out the current image of possible host municipalities to a final disposal facility for spent nuclear fuel, and on the other hand some group interviews of people of another parish but of interest from the municipalities' point of view. The purpose of these group interviews was the same as that of the telephone interview, i.e. to find out what kind of an impact locating a final disposal facility of spent nuclear fuel in a certain municipality would have on the host municipality's image. Because the groups interviewed were selected on different grounds the results of the interviews are not fully comparable. The most important result of the study is that the current attitude towards a final disposal facility for spent nuclear fuel is calm and collected and that the matter is often considered from the standpoint of an outsider. The issue is easily ignored, classified as a matter 'which does not concern me', provided that the facility will not be placed too near one's own home. Among those interviewed the subject seemed not to be of any 'great interest and did not arouse spontaneous feelings for or against'. There are, however, deeply rooted beliefs concerning the facility and quite strong negative and positive attitudes towards it. The facility itself and the associated decision-making procedure arouse many questions, which at present to a large extent are still unexpressed because the subject is considered so remote. It is, however, necessary to give concrete answers to the questions because this makes it possible for people to relate the issue to daily life. It is further important that things arousing fear and doubts also can be discussed because a silence in this respect only emphasizes their importance. The attitude towards the facility is varying. On one hand there are economic and technical factors: the probable economic benefit from it, the obligation to

  15. Proposal of a dry storage installation in Angra NPP for spent nuclear fuel

    International Nuclear Information System (INIS)

    When nuclear fuel is removed from a power reactor core after the depletion of efficiency in generating energy is called Spent Nuclear Fuel (SNF). After its withdrawal from the reactor core, SNF is temporarily stored in pools usually at the same site of the reactor. During this time, short-living radioactive elements and generated heat undergo decay until levels that allow removing the SNF from the pool and sending it for reprocessing or a temporary storage whether any of its final destinations has not yet been defined. It can be loaded in casks and disposed during years in a dry storage installations until be sent to a reprocessing plant or deep repositories. Before any decision, reprocessing or disposal, the SNF needs to be safely and efficiently isolated in one of many types of storages that exist around the world. Worldwide, the amount of SNF increases annually and in the next years this amount will be higher as a consequence of new Nuclear Power Plants (NPP) construction. In Brazil, that is about to construct the Angra 3 nuclear power reactor, a project about the final destination of the SNF is not yet ready. This paper presents a proposal for a dry storage installation in the Angra NPP site since it can be an initial solution for the Brazilian's SNF, until a final decision is taken. (author)

  16. SR 97: post-closure safety for KBS 3 deep repository for spent nuclear fuel - overview -

    International Nuclear Information System (INIS)

    In preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Nuclear Fuel and Waste Management Company, SKB has carried out the long-term safety assessment SR 97, requested by the Swedish Government. The repository is of the KBS-3 type, where the fuel is placed in isolating copper canisters with a high-strength cast iron insert. The canisters are surrounded by bentonite clay in individual deposition holes at a depth of 500 m in granitic bedrock. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock. The future evolution of the repository system is analyzed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings, including climate, persist. The four other scenarios show the evolution if the repository contains a few initially defective canisters, in the event of climate change, in the event of earthquakes, and in the event of future inadvertent human intrusion. The principal conclusion of the assessment is that the prospects of building a safe deep repository for spent nuclear fuel in Swedish granitic bedrock are very good. The results of the assessment also serve as a basis for formulating requirements and preferences regarding the bedrock in site investigations, for designing a program for site investigations, for formulating functional requirements on the repository's barriers, and for prioritization of research. Copyright (2001) Material Research Society

  17. Fusion option to dispose of spent nuclear fuel and transuranic elements

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.

    2000-02-10

    The fusion option is examined to solve the disposition problems of the spent nuclear fuel and the transuranic elements. The analysis of this report shows that the top rated solution, the elimination of the transuranic elements and the long-lived fission products, can be achieved in a fusion reactor. A 167 MW of fusion power from a D-T plasma for sixty years with an availability factor of 0.75 can transmute all the transuranic elements and the long-lived fission products of the 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. The operating time can be reduced to thirty years with use of 334 MW of fusion power, a system study is needed to define the optimum time. In addition, the fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future. Fusion blankets with a liquid carrier for the transuranic elements can achieve a transmutation rate for the transuranic elements up to 80 kg/MW.y of fusion power with k{sub eff} of 0.98. In addition, the liquid blankets have several advantages relative to the other blanket options. The energy from this transmutation is utilized to produce revenue for the system. Molten salt (Flibe) and lithium-lead eutectic are identified as the most promising liquids for this application, both materials are under development for future fusion blanket concepts. The Flibe molten salt with transuranic elements was developed and used successfully as nuclear fuel for the molten salt breeder reactor in the 1960's.

  18. The Decay of Communism: Managing Spent Nuclear Fuel in the Soviet Union, 1937-1991

    Energy Technology Data Exchange (ETDEWEB)

    Hoegselius, Per (History of Science and Technology, Royal Inst. of Technology, Stockholm (Sweden)), e-mail: perho@kth.se

    2010-09-15

    The historical evolution of spent nuclear fuel (SNF) decision-making in Western Europe and North America is already fairly well-known. For the former socialist countries of Eastern Europe, and in particular the Soviet Union, we know less. There have recently been several good studies of Soviet nuclear power history (e.g. Schmid 2004, 2006, Josephson 2005), but none of them has gone into any depth when it comes to SNF, but rather focused on nuclear power reactors, public acceptance, the role of the media, etc. There are also several good overviews available that problematize the radioactive legacy of the Soviet Union, including the SNF and waste issue, but these studies do not address the historical dynamics and evolution of SNF management over a longer period of time; in other words, they fail to explain how and why the present state of affairs have actually come into being. The aim of this paper is to provide historical insight into the dynamics of SNF decision-making in the Soviet Union, from the origins of nuclear engineering in the 1930s to the collapse of the country in 1991. The nuclear fuel system can be described as a large technical system with a variety of interrelated components. The system is 'large' both because it involves key links between geographically disperse activities, and because it involves a variety of technologies, organizations and people that influence the dynamics and evolution of the system. Soviet SNF history is of particular interest in this context, with a nuclear fuel system that was the most complex in the world. The USSR was a pioneer within nuclear power and developed a variety of reactor designs and technologies for uranium mining, conversion and enrichment, as well as for transport, treatment, storage and reprocessing of spent nuclear fuel. It explored both military and civil uses of the atom, and an enormous amount of people and organizations were involved in realizing highly ambitious nuclear programmes. The USSR is

  19. The Swedish Radiation Protection Institute's regulations concerning the final management of spent nuclear fuel and nuclear waste - with background and comments

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-11-01

    This report presents and comments on the Swedish Radiation Protection Institute's Regulations concerning the Protection of Human Health and the Environment in connection with the Final Management of Spent Nuclear Fuel or Nuclear Waste, SSI FS 1998: 1.

  20. Spent nuclear fuel shipping cask handling capabilities of commercial light water reactors

    International Nuclear Information System (INIS)

    This report describes an evaluation of the cask handling capabilities of those reactors which are operating or under construction. A computerized data base that includes cask handling information was developed with information from the literature and utility-supplied data. The capability of each plant to receive and handle existing spent fuel shipping casks was then evaluated. Modal fractions were then calculated based on the results of these evaluations and the quantities of spent fuel projected to be generated by commercial nuclear power plants through 1998. The results indicated that all plants are capable of receiving and handling truck casks. Up to 118 out of 130 reactors (91%) could potentially handle the larger and heavier rail casks if the maximum capability of each facility is utilized. Design and analysis efforts and physical modifications to some plants would be needed to achieve this high rail percentage. These modifications would be needed to satisfy regulatory requirements, increase lifting capabilities, develop rail access, or improve other deficiencies. The remaining 12 reactors were determined to be capable of handling only the smaller truck casks. The percentage of plants that could receive and handle rail casks in the near-term would be reduced to 64%. The primary reason for a plant to be judged incapable of handling rail casks in the near-term was a lack of rail access. The remaining 36% of the plants would be limited to truck shipments. The modal fraction calculations indicated that up to 93% of the spent fuel accumulated by 1998 could be received at federal storage or disposal facilities via rail (based on each plant's maximum capabilities). If the near-term cask handling capabilities are considered, the rail percentage is reduced to 62%

  1. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations

    International Nuclear Information System (INIS)

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual keff of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data

  2. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.

    Energy Technology Data Exchange (ETDEWEB)

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-02-17

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

  3. Calculation Method for the Projection of Future Spent Nuclear Fuel Discharges

    International Nuclear Information System (INIS)

    This report describes the calculation method developed for the projection of future utility spent nuclear fuel (SNF) discharges in regard to their timing, quantity, burnup, and initial enrichment. This projection method complements the utility-supplied RW-859 data on historic discharges and short-term projections of SNF discharges by providing long-term projections that complete the total life cycle of discharges for each of the current U.S. nuclear power reactors. The method was initially developed in mid-1999 to update the SNF discharge projection associated with the 1995 RW-859 utility survey (CRWMS M and O 1996). and was further developed as described in Rev. 00 of this report (CRWMS M and O 2001a). Primary input to the projection of SNF discharges is the utility projection of the next five discharges from each nuclear unit, which is provided via the revised final version of the Energy Information Administration (EIA) 1998 RW-859 utility survey (EIA 2000a). The projection calculation method is implemented via a set of Excel 97 spreadsheets. These calculations provide the interface between receipt of the utility five-discharge projections that are provided in the RW-859 survey, and the delivery of projected life-cycle SNF discharge quantities and characteristics in the format requisite for performing logistics analysis to support design of the Civilian Radioactive Waste Management System (CRWMS). Calculation method improvements described in this report include the addition of a reactor-specific maximum enrichment-based discharge burnup limit. This limit is the consequence of the enrichment limit, currently 5 percent. which is imposed as a Nuclear Regulatory Commission (NRC) license condition on nuclear fuel fabrication plants. In addition, the calculation method now includes the capability for projecting future nuclear plant power upratings, consistent with many such recent plant uprates and the prospect of additional future uprates. Finally. this report

  4. Calculation Method for the Projection of Future Spent Nuclear Fuel Discharges

    Energy Technology Data Exchange (ETDEWEB)

    B. McLeod

    2002-02-28

    This report describes the calculation method developed for the projection of future utility spent nuclear fuel (SNF) discharges in regard to their timing, quantity, burnup, and initial enrichment. This projection method complements the utility-supplied RW-859 data on historic discharges and short-term projections of SNF discharges by providing long-term projections that complete the total life cycle of discharges for each of the current U.S. nuclear power reactors. The method was initially developed in mid-1999 to update the SNF discharge projection associated with the 1995 RW-859 utility survey (CRWMS M&O 1996). and was further developed as described in Rev. 00 of this report (CRWMS M&O 2001a). Primary input to the projection of SNF discharges is the utility projection of the next five discharges from each nuclear unit, which is provided via the revised final version of the Energy Information Administration (EIA) 1998 RW-859 utility survey (EIA 2000a). The projection calculation method is implemented via a set of Excel 97 spreadsheets. These calculations provide the interface between receipt of the utility five-discharge projections that are provided in the RW-859 survey, and the delivery of projected life-cycle SNF discharge quantities and characteristics in the format requisite for performing logistics analysis to support design of the Civilian Radioactive Waste Management System (CRWMS). Calculation method improvements described in this report include the addition of a reactor-specific maximum enrichment-based discharge burnup limit. This limit is the consequence of the enrichment limit, currently 5 percent. which is imposed as a Nuclear Regulatory Commission (NRC) license condition on nuclear fuel fabrication plants. In addition, the calculation method now includes the capability for projecting future nuclear plant power upratings, consistent with many such recent plant uprates and the prospect of additional future uprates. Finally. this report summarizes the

  5. Implementation of Agile project management in spent nuclear fuel characterization process

    International Nuclear Information System (INIS)

    Full text of publication follows. Spent nuclear fuel characterization (SNFC) is a complex process that covers different areas of analysis and whose final goal is to provide an accurate description of spent nuclear fuel (SNF) status for its future classification for storage and transport. The need to reduce the SNFC processing time maintaining the quality of the product has motivated ENUSA to research and implement Agile project management and human performance techniques. The Agile management techniques are focused in accommodate changes or new requirements in the project during the elaboration process without suffering delays or lose of quality. For its SNF projects ENUSA uses 2 complementary techniques: SCRUM and Kanban. SCRUM methodology is based on divide the process into activities blocks. Each block is a finished part of the final product which allows periodical deliveries of the product and the easy introduction of changes if they are necessary. The characterization process is formed by blocks of activities based on different analysis for every fuel assembly as the existence of leaking rods; the analysis of the structural integrity considering the existence of missing rods, broken or missing grids or grid straps or grid springs...; the corrosion phenomenon on the rod that could affect its integrity during the storage and transport; the burnup of the fuel assembly; the analysis of the rod internal pressure and its effect on rod failure mechanism as creep or on the material embrittlement due to the radial hydride precipitation; the compatibility with the container to avoid operational problems during cask loading and unloading, and any new input based on the regulatory evolution and the industry state of the art. The different analysis can be developed at the same time as they are independent. Kanban methodology consists in a visual representation of the evolution of the process. In a chart, the different activities needed to perform any of the analysis

  6. Study of Compton suppression for use in spent nuclear fuel assay

    Science.gov (United States)

    Bender, Sarah

    The focus of this study has been to assess Compton suppressed gamma-ray detection systems for the multivariate analysis of spent nuclear fuel. This objective has been achieved using direct measurement of samples of irradiated fuel elements in two geometrical configurations with Compton suppression systems. In order to address the objective to quantify the number of additionally resolvable photopeaks, direct Compton suppressed spectroscopic measurements of spent nuclear fuel in two configurations were performed: as intact fuel elements and as dissolved feed solutions. These measurements directly assessed and quantified the differences in measured gamma-ray spectrum from the application of Compton suppression. Several irradiated fuel elements of varying cooling time from the Penn State Breazeale Reactor spent fuel inventory were measured using three Compton suppression systems that utilized different primary detectors: HPGe, LaBr3, and NaI(Tl). The application of Compton suppression using a LaBr3 primary detector to the measurement of the current core fuel element, which presented the highest count rate, allowed four additional spectral features to be resolved. In comparison, the HPGe-CSS was able to resolve eight additional photopeaks as compared to the standalone HPGe measurement. Measurements with the NaI(Tl) primary detector were unable to resolve any additional peaks, due to its relatively low resolution. Samples of Approved Test Material (ATM) commercial fuel elements were obtained from Pacific Northwest National Laboratory. The samples had been processed using the beginning stages of the PUREX method and represented the unseparated feed solution from a reprocessing facility. Compton suppressed measurements of the ATM fuel samples were recorded inside the guard detector annulus, to simulate the siphoning of small quantities from the main process stream for long dwell measurement periods. Photopeak losses were observed in the measurements of the dissolved ATM

  7. Investigation of Yucca Mountain repository capacity for the US spent nuclear fuel inventory

    International Nuclear Information System (INIS)

    An analytical decay heat model was developed to evaluate the US inventory of commercial spent nuclear fuel (SNF). The model was benchmarked against the results from ORIGEN-ARP 5.01. The new analytical SNF decay heat model was applied to actual (thru 2002) and projected SNF data. The total decay heat from the 63,000 MT commercial SNF at year 2012 was estimated at 182 MW. According to the thermal loading analysis using a mountain-scale heat transfer model, a 4.9 km2 (1165 acre) site designated for SNF disposal was found to have the capacity to store more SNF than the statutory limit of 70,000 MTIHM. The maximum capacity available for SNF disposal at the Yucca Mountain site is dependent upon the thermal loading strategy chosen and SNF cooling time before emplacement. It was also shown that using high burnup SNF and adjusting the drift spacing, the capacity of the repository could be maximized on a per energy production basis, although some additional cooling may be necessary. Future work needs to consider extending the 'footprint' of the repository, applying non-uniform SNF loading into the drifts, and the impact of spent fuel reprocessing and other decay heat management strategies

  8. Ultrasonic Fingerprinting of Structural Materials: Spent Nuclear Fuel Containers Case-Study

    Science.gov (United States)

    Sednev, D.; Lider, A.; Demyanuk, D.; Kroening, M.; Salchak, Y.

    Nowadays, NDT is mainly focused on safety purposes, but it seems possible to apply those methods to provide national and IAEA safeguards. The containment of spent fuel in storage casks could be dramatically improved in case of development of so-called "smart" spent fuel storage and transfer casks. Such casks would have tamper indicating and monitoring/tracking features integrated directly into the cask design. The microstructure of the containers material as well as of the dedicated weld seam is applied to the lid and the cask body and provides a unique fingerprint of the full container, which can be reproducibly scanned by using an appropriate technique. The echo-sounder technique, which is the most commonly used method for material inspection, was chosen for this project. The main measuring parameter is acoustic noise, reflected from material's artefacts. The purpose is to obtain structural fingerprinting. Reference measurement and additional measurement results were compared. Obtained results have verified the appliance of structural fingerprint and the chosen control method. The successful authentication demonstrates the levels of the feature points' compliance exceeding the given threshold which differs considerably from the percentage of the concurrent points during authentication from other points. Since reproduction or doubling of the proposed unique identification characteristics is impossible at the current state science and technology, application of this technique is considered to identify the interference into the nuclear materials displacement with high accuracy.

  9. Analysis of Spent Fuel Characteristics in Different Scenarios of Closing the Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Calculation analysis of the isotope and radiation-migration characteristics of spent nuclear fuel (SNF) in the open fuel cycle of thermal reactor VVER-1000 and in closed cycle of fast reactor with lead-bismuth coolant has been fulfilled. Effects of including an accelerator-driven system (ADS) into the system for transmutation of minor actinides (MA) into the cycles studied on the SNF characteristics has been reviewed. The application of ADS-burner of MA symbiotically with operating VVER-1000 reactors has been shown to decrease the high-level wastes’ activity approximately 20 times within the interval from the end of cooling in reactor to 105 years; in this case the principle of radiation-migration balance of activity in the underground burial for ~500 years is met as well. The calculation analysis gives grounds to conclude that the use of ADS for burning minor actinides in closed fuel cycle of fast reactors with lead-bismuth coolant, where U, Pu, and MA are recycled, with natural uranium as a makeup fuel, does not provide any special advantages in terms of radiation and migration characteristics of spent fuel and its wastes. (author)

  10. Radiation Exposures Associated with Shipments of Foreign Research Reactor Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Experience has shown that the analyses of marine transport of spent fuel in the Environmental Impact Statement (EIS) were conservative. It is anticipated that for most shipments. The external dose rate for the loaded transportation cask will be more in line with recent shipments. At the radiation levels associated with these shipments, we would not expect any personnel to exceed radiation exposure limits for the public. Package dose rates usually well below the regulatory limits and personnel work practices following ALARA principles are keeping human exposures to minimal levels. However, the potential for Mure shipments with external dose rates closer to the exclusive-use regulatory limit suggests that DOE should continue to provide a means to assure that individual crew members do not receive doses in excess of the public dose limits. As a minimum, the program will monitor cask dose rates and continue to implement administrative procedures that will maintain records of the dose rates associated with each shipment, the vessel used, and the crew list for the vessel. DOE will continue to include a clause in the contract for shipment of the foreign research reactor spent nuclear fuel requiring that the Mitigation Action Plan be followed

  11. Integrated cask storage systems for storage, transportation, and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Since 1979 Tennessee Valley Authority TVA has participated in conceptual design studies of dry storage vaults, silos, casks, ad dry wells, and, with DOE and others, has undertaken limited demonstrations of rod consolidation and cask dry storage at TVA's Browns Ferry Nuclear Plant in Alabama. TVA believes the integrated storage cask concept is worthy of consideration as an alternative for spent fuel management. Placing spent fuel in a secure passive storage mode at an early date and avoiding unnecessary handling and repackaging reduces the potential for occupational and public radiological exposure. Therefore the notion of a universal cask used for storage, shipment, and disposal is appealing from a safety, environmental, and public perception standpoint. The universal cask can also serve as a dispersed monitored retrievable storage (MRS), thus eliminating the need for redundant facilities, and it does not foreclose future options. It also appears that this concept would simplify repository design, ease retrievability, and provide greater flexibility in repository siting. 2 figures, 2 tables

  12. Development of Pyrochemical Reprocessing of the Spent Nuclear Fuel and Prospects of Closed Fuel Cycle

    Directory of Open Access Journals (Sweden)

    R. Tulackova

    2007-01-01

    Full Text Available Molten-Salt Reactor (MSR is a design of an advanced reactor system from the GEN IV family working in thermal or epithermal neutron spectrum and using thorium or transuranium fuel in the form of molten fluorides. It is based on the experience with the development of the molten-salt reactor technology in the Oak-Ridge National Laboratory in the United States. The MSR fuel cycle with integrated reprocessing represents one of the potential ways both for significant decrease of total amount of radioactive wastes for final deposition and for utilization of nuclear energy for electricity and heat production as effectively as possible. There are two pyrochemical reprocessing techniques studied in NRI Rez plc which are considered to be applied both for reprocessing of already existing spent fuel and for preparation and „on-line“ reprocessing of MSR fuel: (i the Fluoride Volatility Method (FVM, which performs chemical conversion of spent thermal oxide fuel components into fluorides and their consequent separation by means of their different volatility, thermal stability and chemical affinity to various sorbents; and (ii electrochemical separation of the actinides (Ans and fission products (FP, represented mainly by lanthanides (Lns, from each other by electrolytic deposition method on solid cathode in molten fluoride media.

  13. Preliminary design report for prototypical spent nuclear fuel rod consolidation equipment

    International Nuclear Information System (INIS)

    The purpose of the Prototypical Consolidation Demonstration Project (PCDP) is to develop and demonstrate the equipment system that will be used to consolidate the bulk of the spent nuclear fuel generated in the United States prior to its placement in a geological repository. The equipment must thus be capable of operating on a routine production basis over a long period of time with stringent requirements for safety, reliability, productivity and cost-effectiveness. Four phases are planned for the PCDP. Phase 1 is the Preliminary Design of generic consolidation equipment that could be installed at a Monitored Retrievable Storage (MRS) facility or in the Receiving ampersand Handling Facility at a geologic repository site. Phase 2 will be the Final Design and preparation of procurement packages for the equipment in a configuration capable of being installed and tested in a special enclosure within the TAN Hot Shop at DOE's Idaho National Engineering Laboratory. In Phase 3 the equipment will be fabricated and then tested with mock fuel elements in a contractor's facility. Finally, in Phase 4 the equipment will be moved to the TAN facility for demonstration operation with irradiated spent fuel elements. 55 figs., 15 tabs

  14. Radiation dosimetry and monitoring for a test of geologic storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    The Lawrence Livermore National Laboratory, at the request of the U.S. Department of Energy, conducted a test at the Nevada Test Site to evaluate both the feasibility of deep data extrapolated to proposed Test and Evaluation Facility and Monitored Retrievable Storage schemes indicate that personnel exposures would be much lower than the yearly guides set by the National Commission on Radiological Protection. The dose commitment in person-rem resulting from spent-fuel storage is extrapolated to be about 0.2% of that currently accepted in the normal operation of nuclear power plants. This report describes the radiological monitoring and dosimetry program that was instituted to limit personnel exposure to ionizing radiation and to determine the extent of any gaseous or particulate release. During the three-year storage phase, no measurable radioactive effluent was released to the atmosphere, soil, or water. Radiation exposures to personnel handling the shielded fuel assemblies (> 50,000 rad/h at contact) were less than 0.4 person-rem for the duration of the project or less than 0.016 person-rem per spent-fuel handling operation

  15. DOCUMENTATION OF NATIONAL WEATHER CONDITIONS AFFECTING LONG-TERM DEGRADATION OF COMMERCIAL SPENT NUCLEAR FUEL AND DOE SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTE

    Energy Technology Data Exchange (ETDEWEB)

    W. L. Poe, Jr.; P.F. Wise

    1998-11-01

    The U.S. Department of Energy (DOE) is preparing a proposal to construct, operate 2nd monitor, and eventually close a repository at Yucca Mountain in Nye County, Nevada, for the geologic disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). As part of this effort, DOE has prepared a viability assessment and an assessment of potential consequences that may exist if the repository is not constructed. The assessment of potential consequences if the repository is not constructed assumes that all SNF and HLW would be left at the generator sites. These include 72 commercial generator sites (three commercial facility pairs--Salem and Hope Creek, Fitzpatrick and Nine Mile Point, and Dresden and Morris--would share common storage due to their close proximity to each other) and five DOE sites across the country. DOE analyzed the environmental consequences of the effects of the continued storage of these materials at these sites in a report titled Continued Storage Analysis Report (CSAR; Reference 1 ) . The CSAR analysis includes a discussion of the degradation of these materials when exposed to the environment. This document describes the environmental parameters that influence the degradation analyzed in the CSAR. These include temperature, relative humidity, precipitation chemistry (pH and chemical composition), annual precipitation rates, annual number of rain-days, and annual freeze/thaw cycles. The document also tabulates weather conditions for each storage site, evaluates the degradation of concrete storage modules and vaults in different regions of the country, and provides a thermal analysis of commercial SNF in storage.

  16. Predicting fissile content of spent nuclear fuel assemblies with the Passive Neutron Albedo Reactivity technique and Monte Carlo code emulation

    International Nuclear Information System (INIS)

    There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed. (author)

  17. Predicting fissile content of spent nuclear fuel assemblies with the passive neutron Albedo reactivity technique and Monte Carlo code emulation

    International Nuclear Information System (INIS)

    There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed.

  18. Spent nuclear fuel as a waste form for geologic disposal: Assessment and recommendations on data and modeling needs

    International Nuclear Information System (INIS)

    This study assesses the status of knowledge pertinent to evaluating the behavior of spent nuclear fuel as a waste form in geologic disposal systems and provides background information that can be used by the DOE to address the information needs that pertain to compliance with applicable standards and regulations. To achieve this objective, applicable federal regulations were reviewed, expected disposal environments were described, the status of spent-fuel modeling was summarized, and information regarding the characteristics and behavior of spent fuel was compiled. This compiled information was then evaluated from a performance modeling perspective to identify further information needs. A number of recommendations were made concerning information still needed to enhance understanding of spent-fuel behavior as a waste form in geologic repositories. 335 refs., 22 figs., 44 tabs

  19. Spent nuclear fuel as a waste form for geologic disposal: Assessment and recommendations on data and modeling needs

    Energy Technology Data Exchange (ETDEWEB)

    Van Luik, A.E.; Apted, M.J.; Bailey, W.J.; Haberman, J.H.; Shade, J.S.; Guenther, R.E.; Serne, R.J.; Gilbert, E.R.; Peters, R.; Williford, R.E.

    1987-09-01

    This study assesses the status of knowledge pertinent to evaluating the behavior of spent nuclear fuel as a waste form in geologic disposal systems and provides background information that can be used by the DOE to address the information needs that pertain to compliance with applicable standards and regulations. To achieve this objective, applicable federal regulations were reviewed, expected disposal environments were described, the status of spent-fuel modeling was summarized, and information regarding the characteristics and behavior of spent fuel was compiled. This compiled information was then evaluated from a performance modeling perspective to identify further information needs. A number of recommendations were made concerning information still needed to enhance understanding of spent-fuel behavior as a waste form in geologic repositories. 335 refs., 22 figs., 44 tabs.

  20. Safeguards for spent nuclear fuel in transfer from wet storage to dry storage in on-site interim storage facilities

    International Nuclear Information System (INIS)

    Germany initially planned to store spent nuclear fuel in the two away-from-reactor interim storage facilities built at Ahaus and Gorleben. The current approach for spent fuel management is on-site interim storage in transport and storage casks as part of a political agreement between the German government and the nuclear operators on the future use of nuclear energy. A reason for this is to avoid near term transportation of spent fuel determined for direct final disposal via public road or rail systems to away-from-reactor storage facilities. Recent legislation has triggered the construction of 12 on-site dry storage facilities at nuclear power plants. Currently, such facilities are being taken into operation on a step-by-step basis. There is a strong need to develop acceptable safeguards concepts for both transfer and dry storage of spent fuel, ideally, a standard safeguards concept that would match all German nuclear power plant sites without ignoring technical and organisational differences. The paper will address the relevant issues and give an overview of the status of safeguards implementation. (author)

  1. Heat pipe cooling system for safer storage of spent fuel in nuclear plant

    International Nuclear Information System (INIS)

    This paper proposes completely passive cooling system utilizing heat pipe thermal diode character and natural air convection cooling of condenser, for cooling spent fuel pool. Detail analysis of various heat pipe design cases to determine the best design concept in terms of cooling power, construction and cost has been presented. The best design when considering thermal safety margin and cost is the heat pipe cooling system with capacity of 0.9 MW, 1,662 heat pipe modules. For this design case, water temperature will reach to peak 68degC after 75 hours, and then it will saturate at 50degC after 2,000 hours. The proposed heat pipe systems can be operated completely passive which will provide safer operational environment to nuclear power plants. (author)

  2. An Empirical Approach to Bounding the Axial Reactivity Effects of PWR Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    P. M. O' Leary; J. M. Scaglione

    2001-04-04

    One of the significant issues yet to be resolved for using burnup credit (BUC) for spent nuclear fuel (SNF) is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters (such as local power, fuel temperature, moderator temperature, burnable poison rod history, and soluble boron concentration) affect the isotopic inventory of fuel that is depleted in a pressurized water reactor (PWR). However, obtaining the detailed operating histories needed to model all PWR fuel assemblies to which BUC would be applied is an onerous and costly task. Simplifications therefore have been suggested that could lead to using ''bounding'' depletion parameters that could be broadly applied to different fuel assemblies. This paper presents a method for determining a set of bounding depletion parameters for use in criticality analyses for SNF.

  3. FY 2016 Status Report: CIRFT Testing on Spent Nuclear Fuels and Hydride Reorientation Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Bevard, Bruce B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

    2016-08-04

    This report provides a detailed description of the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) testing conducted on spent nuclear fuel (SNF) rods in FY 2016, including hydride reorientation test results. Contact-based measurement, or three-LVDT-based curvature measurement, of SNF rods has proven to be quite reliable in CIRFT testing. However, how the linear variable differential transformer (LVDT) head contacts the SNF rod may have a significant effect on the curvature measurement, depending on the magnitude and direction of rod curvature. To correct such contact/curvature issues, sensor spacing, defined as the amount of separation between the three LVDT probes, is a critical measurement that can be used to calculate rod curvature once the deflections are obtained.

  4. Spent Nuclear Fuel (SNF) Cold Vacuum Drying (CVD) Facility Operations Manual

    Energy Technology Data Exchange (ETDEWEB)

    IRWIN, J.J.

    1999-07-02

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-553, Spent Nuclear Fuel Project Final Safety Analysis Report Annex B--Cold Vacuum Drying Facility. The HNF-SD-SNF-DRD-002, 1999, Cold Vacuum Drying Facility Design Requirements, Rev. 4, and the CVDF Final Design Report. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence and references to the CVDF System Design Descriptions (SDDs). This manual has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  5. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    Energy Technology Data Exchange (ETDEWEB)

    IRWIN, J.J.

    2000-02-03

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of the Processing Systems (Garvin 1998) and, the HNF-SD-SNF-DRD-002, 1997, Cold Vacuum Drying Facility Design Requirements, Rev. 3a. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence, and has been developed for the spent nuclear fuel project (SNFP) Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  6. Management Of Hanford KW Basin Knockout Pot Sludge As Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Raymond, R. E. [CH2M HIll Plateau Remediation Company, Richland, WA (United States); Evans, K. M. [AREVA, Avignon (France)

    2012-10-22

    CH2M HILL Plateau Remediation Company (CHPRC) and AREVA Federal Services, LLC (AFS) have been working collaboratively to develop and deploy technologies to remove, transport, and interim store remote-handled sludge from the 10S-K West Reactor Fuel Storage Basin on the U.S. Department of Energy (DOE) Hanford Site near Richland, WA, USA. Two disposal paths exist for the different types of sludge found in the K West (KW) Basin. One path is to be managed as Spent Nuclear Fuel (SNF) with eventual disposal at an SNF at a yet to be licensed repository. The second path will be disposed as remote-handled transuranic (RH-TRU) waste at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, NM. This paper describes the systems developed and executed by the Knockout Pot (KOP) Disposition Subproject for processing and interim storage of the sludge managed as SNF, (i.e., KOP material).

  7. Available reprocessing and recycling services for research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tozser, Sandor; Marshall, Frances M.; Adelfang, Pablo; Bradley, Edward [International Atomic Energy Agency, Vienna (Austria); Budu, Madalina Elena [SOSNY Research and Development Company, Moscow (Russian Federation); Chiguer, Mustapha [AREVA, Paris La Defense (France)

    2016-03-15

    International activities in the back end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. In the future inventories of LEU SNF will continue to be created and the back end solution of RR SNF remains a critical issue. The IAEA, based on the experience gained during the decade of international cooperation in supporting the objectives of the HEU take-back programmes, drew up a report presenting available reprocessing and recycling services for RR SNF. This paper gives an overview of the report, which will address all aspects of reprocessing and recycling services for RR SNF.

  8. Available reprocessing and recycling services for research reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    International activities in the back end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. In the future inventories of LEU SNF will continue to be created and the back end solution of RR SNF remains a critical issue. The IAEA, based on the experience gained during the decade of international cooperation in supporting the objectives of the HEU take-back programmes, drew up a report presenting available reprocessing and recycling services for RR SNF. This paper gives an overview of the report, which will address all aspects of reprocessing and recycling services for RR SNF.

  9. Regulatory standards for permanent disposal of spent nuclear fuel and high-level radioactive waste

    International Nuclear Information System (INIS)

    This paper provides a summary of observations drawn from twenty years of personal experience in working with regulatory criteria for the permanent disposal of radioactive waste for both the Waste Isolation Pilot Plant repository for transuranic defense waste and the proposed Yucca Mountain repository for spent nuclear fuel and high-level wastes. Rather than providing specific recommendations for regulatory criteria, my goal here is to provide a perspective on topics that are fundamental to how high-level radioactive waste disposal regulations have been implemented in the past. What are the main questions raised relevant to long-term disposal regulations? What has proven effective in the past? Where have regulatory requirements perhaps had unintended consequences? New regulations for radioactive waste disposal may prove necessary, but the drafting of these regulations may be premature until a broad range of policy issues are better addressed. In the interim, the perspective offered here may be helpful for framing policy discussions.

  10. A systems evaluation model for selecting spent nuclear fuel storage concepts

    International Nuclear Information System (INIS)

    This paper describes a system evaluation approach used to identify and evaluate monitored, retrievable fuel storage concepts that fulfill ten key criteria for meeting the functional requirements and system objectives of the National Nuclear Waste Management Program. The selection criteria include health and safety, schedules, costs, socio-economic factors and environmental factors. The methodology used to establish the selection criteria, develop a weight of importance for each criterion and assess the relative merit of each storage system is discussed. The impact of cost relative to technical criteria is examined along with experience in obtaining relative merit data and its application in the model. Topics considered include spent fuel storage requirements, functional requirements, preliminary screening, and Monitored Retrievable Storage (MRS) system evaluation. It is concluded that the proposed system evaluation model is universally applicable when many concepts in various stages of design and cost development need to be evaluated

  11. Management of spent nuclear fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee: Environmental assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    On June 1, 1995, DOE issued a Record of Decision [60 Federal Register 28680] for the Department-wide management of spent nuclear fuel (SNF); regionalized storage of SNF by fuel type was selected as the preferred alternative. The proposed action evaluated in this environmental assessment is the management of SNF on the Oak Ridge Reservation (ORR) to implement this preferred alternative of regional storage. SNF would be retrieved from storage, transferred to a hot cell if segregation by fuel type and/or repackaging is required, loaded into casks, and shipped to off-site storage. The proposed action would also include construction and operation of a dry cask SNF storage facility on ORR, in case of inadequate SNF storage. Action is needed to enable DOE to continue operation of the High Flux Isotope Reactor, which generates SNF. This report addresses environmental impacts.

  12. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    International Nuclear Information System (INIS)

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of the Processing Systems (Garvin 1998) and, the HNF-SD-SNF-DRD-002, 1997, Cold Vacuum Drying Facility Design Requirements, Rev. 3a. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence, and has been developed for the spent nuclear fuel project (SNFP) Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  13. Bioremediation of 60Co from simulated spent decontamination solutions of nuclear power reactors by bacteria

    International Nuclear Information System (INIS)

    The spent decontamination solutions generated from nuclear power reactors contain radionuclides of cobalt (60Co, 56Co and 57Co) along with a large excess (105) of non-radioactive metal ions (Fe, Ni and Cr). Our previous studies demonstrated that bioremediation of 60Co from simulated effluents using fungal biomass can provide an alternative to conventional ion exchangers. In this study, we used several bacteria to further improve the process of bioremediation by decreasing biomass requirement and treatment period. Further, metabolite activation in specific bacterial species resulted in enhanced bioremediation of 60Co from simulated effluent. Optimization of conditions in simulated effluent for the eight bacterial species to accomplish maximum 60Co removal is discussed. (author)

  14. Isotopic Analysis of Spent Nuclear Fuel with an Ultra-High Rate HPGe Spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Fast, James E.; Glasgow, Brian D.; Rodriguez, Douglas C.; VanDevender, Brent A.; Wood, Lynn S.

    2014-06-06

    A longstanding challenge is the assay of spent nuclear fuel (SNF). Determining the isotopic content of SNF requires gamma-ray spectroscopy. PNNL has developed new digital filtering and analysis techniques to produce an ultra high-rate gamma-ray spectrometer from a standard coaxial high-purity germanium (HPGe) crystal. This ~40% efficient detector has been operated for SNF measurements at a throughput of about 400k gamma-ray counts per second (kcps) at an input rate of 1.3 Mcps. Optimized filtering algorithms preserve the spectroscopic capability of the system even at these high rates. This talk will present the results of a SNF measurement with aged SNF pellets at PNNL’s Radiochemical Processing Laboratory, first results with a FPGA front end processor capable of processing the data in real time, and the development path toward a multi-element system to assay fuel assemblies.

  15. AIR SHIPMENT OF SPENT NUCLEAR FUEL FROM THE BUDAPEST RESEARCH REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Dewes, J.

    2014-02-24

    The shipment of spent nuclear fuel is usually done by a combination of rail, road or sea, as the high activity of the SNF needs heavy shielding. Air shipment has advantages, e.g. it is much faster than any other shipment and therefore minimizes the transit time as well as attention of the public. Up to now only very few and very special SNF shipments were done by air, as the available container (TUK6) had a very limited capacity. Recently Sosny developed a Type C overpack, the TUK-145/C, compliant with IAEA Standard TS-R-1 for the VPVR/M type Skoda container. The TUK-145/C was first used in Vietnam in July 2013 for a single cask. In October and November 2013 a total of six casks were successfully shipped from Hungary in three air shipments using the TUK-145/C. The present paper describes the details of these shipments and formulates the lessons learned.

  16. Spent nuclear fuel project multi-year work plan WBS number 1.4.1

    International Nuclear Information System (INIS)

    The Spent Nuclear Fuel (SNF) Project Multi-Year Work Plan (MYWP) is a controlled living document that contains the current SNF Project Technical, Schedule and Cost Baselines. These baselines reflect the current Project execution strategies and are controlled via the change control process. Other changes to the MYWP document will be controlled using the document control process. These changes will be processed as they are approved to keep the MYWP a living document. The MYWP will be maintained continuously as the project baseline through the life of the project and not revised annually. The MYWP is the one document which summarizes and links these three baselines in one place. Supporting documentation for each baseline referred to herein may be impacted by changes to the MYWP, and must also be revised through change control to maintain consistency

  17. Designing of a lead ion model source for plasma separation of spent nuclear fuel

    Science.gov (United States)

    Antonov, N. N.; Vorona, N. A.; Gavrikov, A. V.; Samokhin, A. A.; Smirnov, V. P.

    2016-02-01

    Plasma sources of model substances are required for solving problems associated with the development of a plasma separation method for spent nuclear fuel (SNF). Lead is chosen as the substance simulating the kinetics and dynamics of the heavy SNF component. We report on the results of analysis of the discharge in lead vapor with a concentration of 1012-1013 cm-3. Ionization is produced by an electron beam (with electron energy up to 500 eV) in the centimeter gap between planar electrodes. The discharge is simulated using the hydrodynamic and one-particle approximations. The current-voltage characteristics and efficiencies of single ionization depending on the vapor concentrations and thermoelectron current are obtained. The experimentally determined ion currents on the order of 100 μA for an ionization efficiency on the order of 0.1% are in conformity with the result of simulation.

  18. Characterising legacy spent nuclear fuel pond materials using microfocus X-ray absorption spectroscopy.

    Science.gov (United States)

    Bower, W R; Morris, K; Mosselmans, J F W; Thompson, O R; Banford, A W; Law, K; Pattrick, R A D

    2016-11-01

    Analysis of a radioactive, coated concrete core from the decommissioned, spent nuclear fuel cooling pond at the Hunterston-A nuclear site (UK) has provided a unique opportunity to study radionuclides within a real-world system. The core, obtained from a dividing wall and sampled at the fill level of the pond, exhibited radioactivity (dominantly (137)Cs and (90)Sr) heterogeneously distributed across both painted faces. Chemical analysis of the core was undertaken using microfocus spectroscopy at Diamond Light Source, UK. Mapping of Sr across the surface coatings using microfocus X-ray fluorescence (μXRF) combined with X-ray absorption spectroscopy showed that Sr was bound to TiO2 particles in the paint layers, suggesting an association between TiO2 and radiostrontium. Stable Sr and Cs sorption experiments using concrete coupons were also undertaken to assess their interactions with the bulk concrete in case of a breach in the coating layers. μXRF and scanning electron microscopy showed that Sr was immobilized by the cement phases, whilst at the elevated experimental concentrations, Cs was associated with clay minerals in the aggregates. This study provides a crucial insight into poorly understood infrastructural contamination in complex systems and is directly applicable to the UK's nuclear decommissioning efforts. PMID:27262277

  19. COMPLETION OF THE FIRST INTEGRATED SPENT NUCLEAR FUEL TRANSSHIPMENT/INTERIM STORAGE FACILITY IN NW RUSSIA

    International Nuclear Information System (INIS)

    Northwest and Far East Russia contain large quantities of unsecured spent nuclear fuel (SNF) from decommissioned submarines that potentially threaten the fragile environments of the surrounding Arctic and North Pacific regions. The majority of the SNF from the Russian Navy, including that from decommissioned nuclear submarines, is currently stored in on-shore and floating storage facilities. Some of the SNF is damaged and stored in an unstable condition. Existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing this amount of fuel. Additional interim storage capacity is required. Most of the existing storage facilities being used in Northwest Russia do not meet health and safety, and physical security requirements. The United States and Norway are currently providing assistance to the Russian Federation (RF) in developing systems for managing these wastes. If these wastes are not properly managed, they could release significant concentrations of radioactivity to these sensitive environments and could become serious global environmental and physical security issues. There are currently three closely-linked trilateral cooperative projects: development of a prototype dual-purpose transport and storage cask for SNF, a cask transshipment interim storage facility, and a fuel drying and cask de-watering system. The prototype cask has been fabricated, successfully tested, and certified. Serial production is now underway in Russia. In addition, the U.S. and Russia are working together to improve the management strategy for nuclear submarine reactor compartments after SNF removal

  20. A review of the uncertainties in the assessment of radiological consequences of spent nuclear fuel disposal

    International Nuclear Information System (INIS)

    Radioactive waste disposal systems for spent nuclear fuel are designed to isolate the radioactive waste from the human environment for long periods of time. The isolation is provided by a combination of engineered and natural barriers. Safety assessments are performed to describe and quantify the performance of the individual barriers and the disposal system over long-term periods. These assessments will always be associated with uncertainties. Uncertainties can originate from the variability of natural systems and will also be introduced in the predictive modelling performed to quantitatively evaluate the behavior of the disposal system as a consequence of the incomplete knowledge about the governing processes. Uncertainties in safety assessments can partly be reduced by additional measurements and research. The aim of this study has been to identify uncertainties in assessments of radiological consequences from the disposal of spent nuclear fuel based on the Swedish KBS-3 concept. The identified uncertainties have been classified with respect to their origin, i.e. in conceptual, modelling and data uncertainties. The possibilities to reduce the uncertainties are also commented upon. In assessments it is important to decrease uncertainties which are of major importance for the performance of the disposal system. These could to some extent be identified by uncertainty analysis. However, conceptual uncertainties and some types of model uncertainties are difficult to evaluate. To be able to decrease uncertainties in conceptual models, it is essential that the processes describing and influencing the radionuclide transport in the engineered and natural barriers are sufficiently understood. In this study a qualitative approach has been used. The importance of different barriers and processes are indicated by their influence on the release of some representative radionuclides. (au)

  1. Spent Nuclear Fuel Dry Transfer System Cold Demonstration Project Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, Max R; McKinnon, M. A.

    1999-12-01

    The spent nuclear fuel dry transfer system (DTS) provides an interface between large and small casks and between storage-only and transportation casks. It permits decommissioning of reactor pools after shutdown and allows the use of large storage-only casks for temporary onsite storage of spent nuclear fuel irrespective of reactor or fuel handling limitations at a reactor site. A cold demonstration of the DTS prototype was initiated in August 1996 at the Idaho National Engineering and Environmental Laboratory (INEEL). The major components demonstrated included the fuel assembly handling subsystem, the shield plug/lid handling subsystem, the cask interface subsystem, the demonstration control subsystem, a support frame, and a closed circuit television and lighting system. The demonstration included a complete series of DTS operations from source cask receipt and opening through fuel transfer and closure of the receiving cask. The demonstration included both normal operations and recovery from off-normal events. It was designed to challenge the system to determine whether there were any activities that could be made to jeopardize the activities of another function or its safety. All known interlocks were challenged. The equipment ran smoothly and functioned as designed. A few "bugs" were corrected. Prior to completion of the demonstration testing, a number of DTS prototype systems were modified to apply lessons learned to date. Additional testing was performed to validate the modifications. In general, all the equipment worked exceptionally well. The demonstration also helped confirm cost estimates that had been made at several points in the development of the system.

  2. Hanford Spent Nuclear Fuel Project: Recommended path forward. Volume 2: Alternatives and path forward evaluation

    International Nuclear Information System (INIS)

    The Hanford Spent Nuclear Fuel Project has completed an evaluation of four alternatives for expediting the removal of spent nuclear fuel from the K Basins and stabilizing and placing the fuel into interim storage. Four alternatives were compared: (1) Containerizing fuel in the K Basins, transporting fuel to a facility for stabilization, and interim storage of stabilized fuel in a dry storage facility (DSF); (2) Containerizing fuel in the K Basins, transporting fuel to a wet temporary staging facility, moving fuel to a facility for stabilization, and transporting stabilized fuel to an interim DSF; (3) Containerizing fuel in the K Basins in multi-canister overpacks, transporting fuel directly to a stabilization facility for passivation in the overpack, and interim storage of stabilized fuel in a DSF; (4) Packaging fuel for transport overseas and shipping fuel to a foreign reprocessing facility for reprocessing with eventual return of U, Pu and vitrified high level waste. The comparative evaluation consisted of a multi-attribute utility decision analysis, a public, worker and environmental health risk assessment, and a programmatic risk evaluation. The evaluation concluded that the best Path Forward combines the following concepts: Removal of K Basin fuel and sludge is uncoupled from the operation of a stabilization facility; A storage capability is provided to act as a lag storage or staging operation for overpack fuel containers as they are removed from the K Basins; Metal fuel drying and passivation should be maintained as the fuel stabilization process with the option of further refinements as more information becomes available; and The near term NEPA strategy should focus on expeditious removal of fuel and sludge from K Basins and placing overpacked fuel in temporary storage

  3. Evaluation of Nondestructive Assay/Nondestructive Examination Capabilities for Department of Energy Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Luptak, A.J.; Bulmahn, K.D.

    1998-09-01

    This report summarizes an evaluation of the potential use of nondestructive assay (NDA) and nondestructive examination (NDE) technologies on DOE spent nuclear fuel (SNF). It presents the NDA/NDE information necessary for the National Spent Nuclear Fuel Program (NSNFP) and the SNF storage sites to use when defining that role, if any, of NDA/NDE in characterization and certification processes. Note that the potential role for NDA/NDE includes confirmatory testing on a sampling basis and is not restricted to use as a primary, item-specific, data collection method. The evaluation does not attempt to serve as a basis for selecting systems for development or deployment. Information was collected on 27 systems being developed at eight DOE locations. The systems considered are developed to some degree, but are not ready for deployment on the full range of DOE SNF and still require additional development. The system development may only involve demonstrating performance on additional SNF, packaging the system for deployment, and developing calibration standards, or it may be as extensive as performing additional basic research. Development time is considered to range from one to four years. We conclude that NDA/NDE systems are capable of playing a key role in the characterization and certification of DOE SNF, either as the primary data source or as a confirmatory test. NDA/NDE systems will be able to measure seven of the nine key SNF properties and to derive data for the two key properties not measured directly. The anticipated performance goals of these key properties are considered achievable except for enrichment measurements on fuels near 20% enrichment. NDA/NDE systems can likely be developed to measure the standard canisters now being considered for co-disposal of DOE SNF. This ability would allow the preparation of DOE SNF for storage now and the characterization and certification to be finalize later.

  4. DOE Spent Nuclear Fuel Information In Support of TSPA-VA

    Energy Technology Data Exchange (ETDEWEB)

    A. Brewer; D. Cresap; D. Fillmore; H. Loo; M. Ebner; R. McCormack

    1998-09-01

    RW has started the viability assessment (VA) effort to determine the feasibility of Yucca Mountain as the first geologic repository for spent nuclear fuel (SNF) and high-level waste. One component of the viability assessment will be a total system performance assessment (TSPA), based on the design concept and the scientific data and analysis available, describing the repository's probable behavior relative to the overall system performance standards. Thus, all the data collected from the Exploratory Studies Facility to-date have been incorporated into the latest TSPA model. In addition, the Repository Integration Program, an integrated probabilistic simulator, used in the TSPA has also been updated by Golder Associates Incorporated at December 1997. To ensure that the Department of Energy-owned (DOE-owned) SNF continues to be acceptable for disposal in the repository, it will be included in the TSPA-VA evaluation. A number of parameters are needed in the TSPA-VA models to predict the performance of the DOE-owned SNF materials placed into the potential repository. This report documents all of the basis and/or derivation for each of these parameters. A number of properties were not readily available at the time the TSPA-VA data was requested. Thus, expert judgement and opinion was utilized to determine a best property value. The performance of the DOE-owned SNF will be published as part of the TSPA-VA report. Each DOE site will be collecting better data as the DOE SNF program moves closer to repository license application. As required by the RW-0333P, the National Spent Nuclear Fuel Program will be assisting each site in qualifying the information used to support the performance assessment evaluations.

  5. A review of the uncertainties in the assessment of radiological consequences of spent nuclear fuel disposal

    International Nuclear Information System (INIS)

    Radioactive waste disposal systems for spent nuclear fuel are designed to isolate the radioactive waste from the human environment for long period of time. The isolation is provided by a combination of engineered and natural barriers. Safety assessments are performed to describe and quantify the performance of the individual barriers and the disposal system over long-term periods. These assessments will always be associated with uncertainties. Uncertainties can originate from the variability of natural systems and will also be introduced in the predictive modelling performed to quantitatively evaluate the behaviour of the disposal system as a consequence of the incomplete knowledge about the governing processes. Uncertainties in safety assessments can partly be reduced by additional measurements and research. The aim of this study has been to identify uncertainties in assessments of radiological consequences from the disposal of spent nuclear fuel based on the Swedish KBS-3 concept. The identified uncertainties have been classified with respect to their origin, i.e. in conceptual, modelling and data uncertainties. The possibilities to reduce the uncertainties are also commented upon. In assessments it is important to decrease uncertainties which are of major importance for the performance of the disposal system. These could to some extent be identified by uncertainty analysis. However, conceptual uncertainties and some type of model uncertainties are difficult to evaluate. To be able to decrease uncertainties in conceptual models, it is essential that the processes describing and influencing the radionuclide transport in the engineered and natural barriers are sufficiently understood. In this study a qualitative approach has been used. The importance of different barriers and processes are indicated by their influence on the release of some representative radionuclides. (122 refs.) (au)

  6. Safety case for the disposal of spent nuclear fuel at Olkiluoto. Features, events and processes 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-12-15

    Features, Events and Processes sits within Posiva Oy's Safety Case 'TURVA-2012' portfolio and has the objective of presenting the main features, events and processes (FEPs) that are considered to be potentially significant for the long-term safety of the planned KBS-3V repository for spent nuclear fuel at Olkiluoto. The primary purpose of this report is to support Performance Assessment, Formulation of Radionuclide Release Scenarios, Assessment of the Radionuclide Release Scenarios for the Repository System and Biosphere Assessment by ensuring that the scenarios are comprehensive and take account of all significant FEPs. The main FEPs potentially affecting the disposal system are described for each relevant subsystem component or barrier (i.e. the spent nuclear fuel, the canister, the buffer and tunnel backfill, the auxiliary components, the geosphere and the surface environment). In addition, a small number of external FEPs that may potentially influence the evolution of the disposal system are described. The conceptual understanding and operation of each FEP is described, together with the main features (variables) of the disposal system that may affect its occurrence or significance. Olkiluoto-specific issues are considered when relevant. The main uncertainties (conceptual and parameter/data) associated with each FEP that may affect understanding are also documented. Indicative parameter values are provided, in some cases, to illustrate the magnitude or rate of a process, but it is not the intention of this report to provide the complete set of numerical values that are used in the quantitative safety assessment calculations. Many of the FEPs are interdependent and, therefore, the descriptions also identify the most important direct couplings between the FEPs. This information is used in the formulation of scenarios to ensure the conceptual models and calculational cases are both comprehensive and representative. (orig.)

  7. Deployment Evaluation Methodology for the Electrometallurgical Treatment of DOE-EM Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    The Department of Energy - Environmental Management (DOE-EM) National Spent Nuclear Fuel Program (NSNFP) is charged with the disposition of legacy spent nuclear fuel (SNF). While direct repository disposal of the SNF is the preferred disposition option, some DOE SNF may need treatment to meet acceptance criteria at various disposition sites. The treatments may range from electrometallurgical treatment (EMT) and chemical dissolution to engineering controls. As a planning basis, a need is assumed for a treatment process, either as a primary or backup technology, that is compatible with, and cost-effective for, this portion of the DOE-EM inventory. The current planning option for treating this SNF, pending completion of development work and National Environmental Policy Act (NEPA) analysis, is the EMT process under development by Argonne National Laboratory - West (ANL-W). A decision on the deployment of the EMT is pending completion of an engineering scale demonstration currently in progress at ANL-W. For this study, a set of questions was developed for the EMT process for fuels at several locations. The set of questions addresses all issues associated with design, construction, and operation of a production facility. A matrix table was developed to determine questions applicable to various fuel treatment options. A work breakdown structure (WBS) was developed to identify a treatment process and costs from initial design to shipment of treatment products to final disposition. Costs were applied to determine the life-cycle cost of each option. This technique can also be applied to other treatment techniques for treating SNF

  8. Spent nuclear fuel retrieval system fuel handling development testing. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, D.R.; Meeuwsen, P.V.

    1997-09-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin, clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge), remove the contents from the canisters and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. This report describes fuel handling development testing performed from May 1, 1997 through the end of August 1997. Testing during this period was mainly focused on performance of a Schilling Robotic Systems` Conan manipulator used to simulate a custom designed version, labeled Konan, being fabricated for K-Basin deployment. In addition to the manipulator, the camera viewing system, process table layout, and fuel handling processes were evaluated. The Conan test manipulator was installed and fully functional for testing in early 1997. Formal testing began May 1. The purposes of fuel handling development testing were to provide proof of concept and criteria, optimize equipment layout, initialize the process definition, and identify special needs/tools and required design changes to support development of the performance specification. The test program was set up to accomplish these objectives through cold (non-radiological) development testing using simulated and prototype equipment.

  9. Safety case for the disposal of spent nuclear fuel at Olkiluoto. Features, events and processes 2012

    International Nuclear Information System (INIS)

    Features, Events and Processes sits within Posiva Oy's Safety Case 'TURVA-2012' portfolio and has the objective of presenting the main features, events and processes (FEPs) that are considered to be potentially significant for the long-term safety of the planned KBS-3V repository for spent nuclear fuel at Olkiluoto. The primary purpose of this report is to support Performance Assessment, Formulation of Radionuclide Release Scenarios, Assessment of the Radionuclide Release Scenarios for the Repository System and Biosphere Assessment by ensuring that the scenarios are comprehensive and take account of all significant FEPs. The main FEPs potentially affecting the disposal system are described for each relevant subsystem component or barrier (i.e. the spent nuclear fuel, the canister, the buffer and tunnel backfill, the auxiliary components, the geosphere and the surface environment). In addition, a small number of external FEPs that may potentially influence the evolution of the disposal system are described. The conceptual understanding and operation of each FEP is described, together with the main features (variables) of the disposal system that may affect its occurrence or significance. Olkiluoto-specific issues are considered when relevant. The main uncertainties (conceptual and parameter/data) associated with each FEP that may affect understanding are also documented. Indicative parameter values are provided, in some cases, to illustrate the magnitude or rate of a process, but it is not the intention of this report to provide the complete set of numerical values that are used in the quantitative safety assessment calculations. Many of the FEPs are interdependent and, therefore, the descriptions also identify the most important direct couplings between the FEPs. This information is used in the formulation of scenarios to ensure the conceptual models and calculational cases are both comprehensive and representative. (orig.)

  10. Comparative analysis of radiation characteristics from various types of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kryuchkov, E.F.; Opalovsky, V.A.; Tikhomirov, G.V. [Moscow Engineering Physics Institute (State University) (Russian Federation)

    2003-07-01

    At the present time, in purposes of the most effective utilization of nuclear materials, new advanced fuel cycles are under development. These cycles imply application of uranium-plutonium, uranium-thorium and some other types of nuclear fuel. However, it is obvious that the parameters of new nuclear fuel (NF) types will be quite different from those for traditional NF types. These differences can affect significantly the conditions for storage, transportation and reprocessing of spent nuclear fuel (SNF). So, it is necessary to carry out a comparative analysis of radiation characteristics for various NF types at different stages of nuclear fuel cycle (NFC). The present paper addresses radiation properties of the following NF types: UO{sub 2}, UO{sub 2}-PuO{sub 2}, ThO{sub 2}-PaO{sub 2}-UO{sub 2}. Numerical studies have been carried out to determine radiation properties of these NF types at the following NFC stages: radiation properties of NF directly before and after irradiation in the reactor core, after different cooling time, radiation properties of uranium and plutonium fractions after chemical separation, radiation properties of NF re-fabricated for recycle, radiation properties of NF after the second and third recycles. The computer code package SCALE is used for evaluating the radiation properties of different SNF types. Finally, the following major conclusions can be made: 1) Correct description of SNF radiation and dosimetric properties requires available benchmark data on contents of heavy nuclides in SNF; 2) ThO{sub 2}-PaO{sub 2}-UO{sub 2} fuel demonstrates an important feature: internal transmutation of minor actinides provided the ultra-high fuel burn-up is achieved.

  11. Spent fuel management strategy for future nuclear power plants operation in Indonesia

    International Nuclear Information System (INIS)

    Full text: The demand for electricity in Indonesia increases by the years. This increase goes along with the rate of economic development, the rate of population growth and the rapid development in the industrial sector. To fulfill this demand for electricity, it is becoming more difficult to depend on the existing resources which are now getting limited. It is, therefore, very important. that steps should be taken to seek outer resources as alternatives. Based on the thought that a Nuclear Power Plant (NPP) is technically safe, reliable, clean and environmentally-oriented, relatively economical, and supported by our being prepared in respect to the human resources and the infrastructures, including the results of the feasibility studies for NPP development completed in 1996, the option of nuclear power could well be the right solution. Intensive effort have long been undertaken by BATAN for the introduction of NPPs in Indonesia. The candidate sites, which have long been thoroughly investigated and selected, are located at the Muria Peninsula in Central Java. It is planned that the selected candidate site will accommodate several NPPs with a total power generation of about 7,000 MWe. Indonesia shall choose an open fuel cycle strategy until couple decades after the first NPP operation. Each NPP to be capable of providing 3-5 years of spent fuel storage pool at the NPP building. It is planned that the spent fuels will be stored for further 40-50 years in centralized storage facility at the NPP site. After that, spent-fuels will be stored in long term storage facility, eventually, arrangement must be made for deed geological disposal. According to the result of NPP feasibility studies, for achieving 7,000 MWe power generation there are three cases of scenario among many combination of 600 MWe and 900 MWe class NPPs: - Case I gives the power combination of 600 MWe x 2 Units x 6 Stages. - Case 2 gives the combination of following power: 600 MWe x 2 Units x 3 Stages 900

  12. Spent Fuel Test-Climax: An evaluation of the technical feasibility of geologic storage of spent nuclear fuel in granite: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Patrick, W.C. (comp.)

    1986-03-30

    In the Climax stock granite on the Nevada Test Site, eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized. When test data indicated that the test objectives were met during the 3-year storage phase, the spent-fuel canisters were retrieved and the thermal sources were de-energized. The project demonstrated the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner. In addition to emplacement and retrieval operations, three exchanges of spent-fuel assemblies between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. The test led to development of a technical measurements program. To meet these objectives, nearly 1000 instruments and a computer-based data acquisition system were deployed. Geotechnical, seismological, and test status data were recorded on a continuing basis for the three-year storage phase and six-month monitored cool-down of the test. This report summarizes the engineering and scientific endeavors which led to successful design and execution of the test. The design, fabrication, and construction of all facilities and handling systems are discussed, in the context of test objectives and a safety assessment. The discussion progresses from site characterization and experiment design through data acquisition and analysis of test data in the context of design calculations. 117 refs., 52 figs., 81 tabs.

  13. Spent Fuel Test-Climax: An evaluation of the technical feasibility of geologic storage of spent nuclear fuel in granite: Final report

    International Nuclear Information System (INIS)

    In the Climax stock granite on the Nevada Test Site, eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized. When test data indicated that the test objectives were met during the 3-year storage phase, the spent-fuel canisters were retrieved and the thermal sources were de-energized. The project demonstrated the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner. In addition to emplacement and retrieval operations, three exchanges of spent-fuel assemblies between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. The test led to development of a technical measurements program. To meet these objectives, nearly 1000 instruments and a computer-based data acquisition system were deployed. Geotechnical, seismological, and test status data were recorded on a continuing basis for the three-year storage phase and six-month monitored cool-down of the test. This report summarizes the engineering and scientific endeavors which led to successful design and execution of the test. The design, fabrication, and construction of all facilities and handling systems are discussed, in the context of test objectives and a safety assessment. The discussion progresses from site characterization and experiment design through data acquisition and analysis of test data in the context of design calculations. 117 refs., 52 figs., 81 tabs

  14. Long-term lining performance - Civil engineering problem of potential retrieval of buried spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vasicek, Radek, E-mail: radek.vasicek@fsv.cvut.c [Czech Technical University in Prague, Faculty of Civil Engineering, Centre of Experimental Geotechnics, Thakurova 7, 16629 Prague 6 (Czech Republic); Svoboda, Jiri, E-mail: jiri.svoboda@seznam.c [Czech Technical University in Prague, Faculty of Civil Engineering, Centre of Experimental Geotechnics, Thakurova 7, 16629 Prague 6 (Czech Republic)

    2011-04-15

    The current solution for the spent fuel, high-level and long-lived radioactive waste is to store them at surface facilities from which they will be subsequently moved to a deep repository. No such repositories are in operation currently but several such facilities are close to the construction phase. A deep repository can be situated in several types of geological conditions including clay formations, salt sediments, argillites and tuffitic and granitic rocks. The character of the host rock is the key factor determining the design and specific requirements of individual components of such a facility. The future potential retrieval of canisters containing nuclear waste from the repository is a further influential factor. The reason for retrieval of containers lies in the development of fast reactors and increased interest for spent fuel reprocessing. Naturally, the decision as to whether retrievability is technically feasible must be made before finalising the design and construction process of the repository. If the decision is made to retrieve, a design which will include all the relevant safety aspects for the potential retrieval of canisters must be determined. The lay-out of the repository, the materials to be used and the design of the various structures of the facility (e.g. access tunnels, disposal shafts, buffer and backfill) are not the only issues to be addressed. The long-term stability of the system as a whole, i.e. of all the components, is crucial. Depending on the disposal concept chosen, the thermal load generated by the waste in the disposal container, saturation by water from the surrounding environment and the loading of the host rock massif will constitute the main processes which will affect the behaviour, safety and future functioning of the repository from the civil engineering point of view. The long-term stability of the lining of disposal galleries is a basic precondition for the safe removal of spent nuclear waste from deep underground

  15. Draft environmental impact statement on a proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Volume 2, Appendix F, Description and impacts of storage technology alternatives

    International Nuclear Information System (INIS)

    This appendix presents a description and evaluation of currently available spent nuclear fuel storage technologies, and their applicability to foreign research reactor spent nuclear fuel. These technologies represent the range of alternatives dW would be available to implement the proposed action. Some of these technologies are currently in use at US Department of Energy (DOE) facilities. Several dry storage cask and/or building designs have been licensed by the US Nuclear Regulatory Commission (NRC) and are operational with commercial nuclear power plant spent fuel at several locations. This appendix also discusses potential storage sites and impacts of foreign research reactor spent nuclear fuel storage at these locations

  16. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    International Nuclear Information System (INIS)

    This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors

  17. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Y.C. [Square Y, Orchard Park, NY (United States); Chen, S.Y.; LePoire, D.J. [Argonne National Lab., IL (United States). Environmental Assessment and Information Sciences Div.; Rothman, R. [USDOE Idaho Field Office, Idaho Falls, ID (United States)

    1993-02-01

    This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors.

  18. REGIONAL BINNING FOR CONTINUED STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTES

    International Nuclear Information System (INIS)

    In the Continued Storage Analysis Report (CSAR) (Reference 1), DOE decided to analyze the environmental consequences of continuing to store the commercial spent nuclear fuel (SNF) at 72 commercial nuclear power sites and DOE-owned spent nuclear fuel and high-level waste at five Department of Energy sites by region rather than by individual site. This analysis assumes that three commercial facilities pairs--Salem and Hope Creek, Fitzpatrick and Nine-Mile Point, and Dresden and Moms--share common storage due to their proximity to each other. The five regions selected for this analysis are shown on Figure 1. Regions 1, 2, and 3 are the same as those used by the Nuclear Regulatory Commission in their regulatory oversight of commercial power reactors. NRC Region 4 was subdivided into two regions to more appropriately define the two different climates that exist in NRC Region 4. A single hypothetical site in each region was assumed to store all the SNF and HLW in that region. Such a site does not exist and has no geographic location but is a mathematical construct for analytical purposes. To ensure that the calculated results for the regional analyses reflect appropriate inventory, facility and material degradation, and radionuclide transport, the waste inventories, engineered barriers, and environmental conditions for the hypothetical sites were developed from data for each of the existing sites within the given region. Weighting criteria to account for the amount and types of SNF and HLW at each site were used in the development of the environmental data for the regional site, such that the results of the analyses for the hypothetical site were representative of the sum of the results of each actual site if they had been modeled independently. This report defines the actual site data used in development of this hypothetical site, shows how the individual site data was weighted to develop the regional site, and provides the weighted data used in the CSAR analysis. It is

  19. REGIONAL BINNING FOR CONTINUED STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTES

    Energy Technology Data Exchange (ETDEWEB)

    W. Lee Poe, Jr

    1998-10-01

    In the Continued Storage Analysis Report (CSAR) (Reference 1), DOE decided to analyze the environmental consequences of continuing to store the commercial spent nuclear fuel (SNF) at 72 commercial nuclear power sites and DOE-owned spent nuclear fuel and high-level waste at five Department of Energy sites by region rather than by individual site. This analysis assumes that three commercial facilities pairs--Salem and Hope Creek, Fitzpatrick and Nine-Mile Point, and Dresden and Moms--share common storage due to their proximity to each other. The five regions selected for this analysis are shown on Figure 1. Regions 1, 2, and 3 are the same as those used by the Nuclear Regulatory Commission in their regulatory oversight of commercial power reactors. NRC Region 4 was subdivided into two regions to more appropriately define the two different climates that exist in NRC Region 4. A single hypothetical site in each region was assumed to store all the SNF and HLW in that region. Such a site does not exist and has no geographic location but is a mathematical construct for analytical purposes. To ensure that the calculated results for the regional analyses reflect appropriate inventory, facility and material degradation, and radionuclide transport, the waste inventories, engineered barriers, and environmental conditions for the hypothetical sites were developed from data for each of the existing sites within the given region. Weighting criteria to account for the amount and types of SNF and HLW at each site were used in the development of the environmental data for the regional site, such that the results of the analyses for the hypothetical site were representative of the sum of the results of each actual site if they had been modeled independently. This report defines the actual site data used in development of this hypothetical site, shows how the individual site data was weighted to develop the regional site, and provides the weighted data used in the CSAR analysis. It is

  20. Source study for a concept of spent nuclear fuel and radioactive waste management in the Czech Republic

    International Nuclear Information System (INIS)

    The sections of the study are as follows: Purpose of the study; Basic principles of peaceful uses of nuclear energy and spent nuclear fuel and radioactive waste management; Basic starting facts and assumptions and assessment of the fulfilment of the targets; Low-level and medium-level radioactive waste management; Management of spent nuclear fuels and wastes not admitted to near-surface repositories; Description of / proposal for the methodology of transparency policy and proposal for a method to involve the entities affected, including municipalities and the public; Current financial arrangements (reserves for decommissioning, nuclear account); Refinement/modification of the plans and technical solutions; Proposed schedule and milestones for attaining the targets; Proposal for a research and development programme; Assessment of the costs of implementation of the targets. (P.A.)

  1. DEVELOPMENT OF THE SWEDISH DEEP REPOSITORY FOR SPENT NUCLEAR FUEL IN CRYSTALLINE HOST ROCK

    International Nuclear Information System (INIS)

    The Swedish Nuclear Fuel and Waste Management Company, SKB, has developed a system that ensures the safe handling of all kinds of radioactive waste from the Swedish nuclear power plants for a long time period ahead. The keystones of this system are: A transport system with the ship M/S Sigyn which has been in operation since 1983. A central interim storage facility for spent nuclear fuel, CLAB, in operation since 1985. A final repository for short-lived, low and intermediate level waste, SFR, in operation since 1988. In Sweden, the preferred method for final disposal of spent fuel is to encapsulate it in copper canisters and dispose them in a deep geological repository in crystalline host rock. SKB is planning to build an encapsulation plant adjacent to the central storage for spent fuel, CLAB. The siting for the deep repository has not yet been selected. A siting program with feasibility studies was completed in 2001. Early 2002 SKB received the necessary permits to start the site investigation at two potential sites for siting of the deep repository in Sweden. The site investigation at these sites started early 2002 and will be completed during 2007. Over the years, a number of generic studies of the layout of the operational area(s) above ground and underground facilities have been performed. During the site investigation phase the deep repository will be developed to conceptual design status and a number of design studies will be performed. These design studies are called Design Justification Statements (DJS). One important DJS is the selection of access routes from the ground level to the disposal level at tentatively 500 m depth and that study will be completed shortly. The repository design and layout of the disposal areas will be based on site specific conditions and results from demonstration of handling and equipment for canisters, buffer and backfilling. Some of these demonstrations have already been performed at Dspv HRL but additional development and

  2. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement

    International Nuclear Information System (INIS)

    The US Department of Energy (DOE) is engaged in two related decision making processes concerning: (1) the transportation, receipt, processing, and storage of spent nuclear fuel (SNF) at the DOE Idaho National Engineering Laboratory (INEL) which will focus on the next 10 years; and (2) programmatic decisions on future spent nuclear fuel management which will emphasize the next 40 years. DOE is analyzing the environmental consequences of these spent nuclear fuel management actions in this two-volume Environmental Impact Statement (EIS). Volume 1 supports broad programmatic decisions that will have applicability across the DOE complex and describes in detail the purpose and need for this DOE action. Volume 2 is specific to actions at the INEL. This document, which limits its discussion to the Savannah River Site (SRS) spent nuclear fuel management program, supports Volume 1 of the EIS. Following the introduction, Chapter 2 contains background information related to the SRS and the framework of environmental regulations pertinent to spent nuclear fuel management. Chapter 3 identifies spent nuclear fuel management alternatives that DOE could implement at the SRS, and summarizes their potential environmental consequences. Chapter 4 describes the existing environmental resources of the SRS that spent nuclear fuel activities could affect. Chapter 5 analyzes in detail the environmental consequences of each spent nuclear fuel management alternative and describes cumulative impacts. The chapter also contains information on unavoidable adverse impacts, commitment of resources, short-term use of the environment and mitigation measures

  3. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs, Draft Environmental Impact Statement

    International Nuclear Information System (INIS)

    Volume 1 to the Department of Energy's Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Management Programs Environmental Impact Statement evaluates a range of alternatives for managing naval spent nuclear fuel expected to be removed from US Navy nuclear-powered vessels and prototype reactors through the year 2035. The Environmental Impact Statement (EIS) considers a range of alternatives for examining and storing naval spent nuclear fuel, including alternatives that terminate examination and involve storage close to the refueling or defueling site. The EIS covers the potential environmental impacts of each alternative, as well as cost impacts and impacts to the Naval Nuclear Propulsion Program mission. This Appendix covers aspects of the alternatives that involve managing naval spent nuclear fuel at four naval shipyards and the Naval Nuclear Propulsion Program Kesselring Site in West Milton, New York. This Appendix also covers the impacts of alternatives that involve examining naval spent nuclear fuel at the Expended Core Facility in Idaho and the potential impacts of constructing and operating an inspection facility at any of the Department of Energy (DOE) facilities considered in the EIS. This Appendix also considers the impacts of the alternative involving limited spent nuclear fuel examinations at Puget Sound Naval Shipyard. This Appendix does not address the impacts associated with storing naval spent nuclear fuel after it has been inspected and transferred to DOE facilities. These impacts are addressed in separate appendices for each DOE site

  4. 77 FR 4807 - Revised Fee Policy for Acceptance of Foreign Research Reactor Spent Nuclear Fuel From High-Income...

    Science.gov (United States)

    2012-01-31

    ... circumstances. DOE also reserves the right to adjust the fee set in an acceptance contract if there are unique... Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel'' (61 FR 25092, May... FR 26507, May 28, 1996). The policy was subsequently revised only to address the question of...

  5. Spent Nuclear Fuel Project path forward: nuclear safety equivalency to comparable NRC-licensed facilities

    International Nuclear Information System (INIS)

    This document includes the Technical requirements which meet the nuclear safety objectives of the NRC regulations for fuel treatment and storage facilities. These include requirements regarding radiation exposure limits, safety analysis, design and construction. This document also includes administrative requirements which meet the objectives of the major elements of the NRC licensing process. These include formally documented design and safety analysis, independent technical review, and oppportunity for public involvement

  6. Spent Nuclear Fuel Project path forward: nuclear safety equivalency to comparable NRC-licensed facilities

    Energy Technology Data Exchange (ETDEWEB)

    Garvin, L.J.

    1995-11-01

    This document includes the Technical requirements which meet the nuclear safety objectives of the NRC regulations for fuel treatment and storage facilities. These include requirements regarding radiation exposure limits, safety analysis, design and construction. This document also includes administrative requirements which meet the objectives of the major elements of the NRC licensing process. These include formally documented design and safety analysis, independent technical review, and oppportunity for public involvement.

  7. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package

  8. TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

    International Nuclear Information System (INIS)

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. Fifty-seven UO2, UO2/Gd2O3, and UO2/PuO2 critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on keff (which can be a function of the trending parameters) such that the biased keff, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading

  9. Attitudes towards a final repository for the spent nuclear fuel. Structure and causes

    International Nuclear Information System (INIS)

    This report presents the results of a questionnaire survey of attitudes towards a final repository for the spent nuclear fuel. The questionnaire was mailed to 3,000 persons. Participants were young and older people in Oskarshamn municipality and Oesthammar municipality as well as in the rest of the country. Fifty-one percent responded. The questionnaire included a large number of questions of possible relevance for understanding the structure of and reasons for the person's attitude towards a final repository. Questions concerning nuclear power were dealt with in a special section. Men were more positively disposed towards a repository than women, older people more than young. The gender differences are mainly attributable to the variation in attitude towards nuclear power and concern about nuclear accidents. In the case of older people, interest was also a factor. Young people were not as interested in the issue. The most important factor in determining the attitude towards a final repository was the benefit it was expected to bring to the municipality. Moral and emotional aspects were also important. Risk played a relatively subordinate role. Social aspects were very important: those who frequently spoke with people who were positively disposed tended to be positive themselves, and vice versa for those who were negative. This factor could explain some of the gender differences in attitude. Attitudes in Oskarshamn were slightly more positive than in Oesthammar, probably due to the fact that the residents in Oskarshamn had a greater sense of participation in the municipality's decision in the matter. Information from SKB was also found to be an important factor for the differences in attitude between the municipalities. Eight percentage points more people had received information in Oskarshamn than in Oesthammar. The difference may be small, but it exists and does appear to be of some importance. Attitudes in Oskarshamn and Oesthammar continued to be much more

  10. Integrated data base report--1996: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and commercial and U.S. government-owned radioactive wastes. Inventories of most of these materials are reported as of the end of fiscal year (FY) 1996, which is September 30, 1996. Commercial SNF and commercial uranium mill tailings inventories are reported on an end-of-calendar year (CY) basis. All SNF and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are SNF, high-level waste, transuranic waste, low-level waste, uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, naturally occurring and accelerator-produced radioactive material, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through FY 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

  11. Integrated data base report - 1996: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and commercial and U.S. government-owned radioactive wastes. Inventories of most of these materials are reported as of the end of fiscal year (FY) 1996, which is September 30, 1996. Commercial SNF and commercial uranium mill tailings inventories are reported on an end-of-calendar year (CY) basis. All SNF and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are SNF, high-level waste, transuranic waste, low-level waste, uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, naturally occurring and accelerator-produced radioactive material, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through FY 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions

  12. The geological storage of spent nuclear fuel and depleted uranium beneath the Williston Basin

    Energy Technology Data Exchange (ETDEWEB)

    Brunskill, B. [Helix Geological Consultants, Regina, SK (Canada)

    2007-07-01

    In order to prevent or retard the leakage of buried nuclear material into the surrounding rocks, regulatory agencies in Canada and the United States are recommending that spent nuclear fuel eventually be stored in suitable geological repositories with highly-engineered barriers. This presentation discussed the development of a repository somewhere in the Precambrian Shield beneath the Williston Basin in Canada, as well as a repository that was under construction at Yucca Mountain, Nevada, in the United States. Potential storage sites in Canada were provided in an illustration and a figure of the proposed repository development was provided. Other illustrations included a light-water fuel rod and assembly as well as storage containers and drip shield. It was shown that in order to prevent potential migration, it would be highly beneficial if a repository were located where the groundwater surrounding the repository was not vertically mobile. A map of the Williston Basin boundary and a fluid-flow model and alternate model through the Williston Basin were also presented. The primary benefits of developing a deep geological repository were presented. These included a favourable hydrogeological regime which would likely isolate and contain the eventual release of any radioactive material. Other benefits that were discussed included minimal disturbance to the geological media during development; elimination of most underground-related mining construction; and, radiation safety issues. tabs., figs.

  13. ARTIST process. A novel chemical process for treatment of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tachimori, Shoichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-10-01

    A new chemical process, ARTIST process, is proposed for the treatment of spent nuclear fuel. The main concept of the ARTIST process is to recover and stock all actinides (Ans) as two groups, uranium (U) and a mixture of transuranics (TRU), to preserve their resource value and to dispose solely fission products (FPs). The process is composed of two main steps, an U exclusive isolation and a total recovery of TRU; which copes with the nuclear non-proliferation measures, and additionally of Pu separation process and soft N-donor process if requested, and optionally of processes for separation of long-lived FPs. These An products: U-product and TRU-product, are to be solidified by calcination and allowed to the interim stockpile for future utilization. These separations are achieved by use of amidic extractants in accord with the CHON principle. The technical feasibility of the ARTIST process was explained by the performance of both the branched alkyl monoamides in extracting U and suppressing the extraction of tetravalent Ans due to the steric effect and the diglycolic amide (TODGA) in thorough extraction of all TRU by tridentate fashion. When these TRU are requested to put into reactors, LWR or FBR, for power generation or the Accelerator - Driven System (ADS) for transmutation, Pu (Np) or Am-Cm (Np) are to be extracted from the TRU-product. (author)

  14. Issues relating to spent nuclear fuel storage on the Oak Ridge Reservation

    International Nuclear Information System (INIS)

    Currently, about 2,800 metric tons of spent nuclear fuel (SNF) is stored in the US, 1,000 kg of SNF (or about 0.03% of the nation's total) are stored at the US Department of Energy (DOE) complex in Oak Ridge, Tennessee. However small the total quantity of material stored at Oak Ridge, some of the material is quite singular in character and, thus, poses unique management concerns. The various types of SNF stored at Oak Ridge will be discussed including: (1) High-Flux Isotope Reactor (HFIR) and future Advanced Neutron Source (ANS) fuels; (2) Material Testing Reactor (MTR) fuels, including Bulk Shielding Reactor (BSR) and Oak Ridge Research Reactor (ORR) fuels; (3) Molten Salt Reactor Experiment (MSRE) fuel; (4) Homogeneous Reactor Experiment (HRE) fuel; (5) Miscellaneous SNF stored in Oak Ridge National Laboratory's (ORNL's) Solid Waste Storage Areas (SWSAs); (6) SNF stored in the Y-12 Plant 9720-5 Warehouse including Health. Physics Reactor (HPRR), Space Nuclear Auxiliary Power (SNAP-) 10A, and DOE Demonstration Reactor fuels

  15. Technology Development for Integrated Safety Test of Spent Nuclear Fuel Transportation and Storage System

    International Nuclear Information System (INIS)

    A dedicated review on the U. S. NRC Regulation 10 CFR Part 72 and regulatory guide NUREG/1536 has been performed. The safety requirements for spent nuclear fuel dry storage cask are analyzed and summarized in structural, thermal, shielding, criticality, materials, tests and maintenance aspects. Also a guideline for preparing the safety analysis report is provided. The heat flow analysis was performed by varying the dimensions of the heat flow test facility. From the heat flow analysis for the test facility, as the test facility became test facility. From the heat flow analysis for the test facility, as the test facility became bigger; the thermal effect became smaller. Therefore, the dimensions of the heat flow test facility was designed with 5m Χ 5m Χ 6m(H). Analyses of heat transfer characteristics and mechanism for spent PWR fuel assemblies, option study for production of the effective thermal conductivity and option study for effective thermal conductivity test have been performed to obtain the basic data for production of the effective thermal conductivity. It became clear that the diffusion coefficient of chloride ion of concrete remarkably increases along with the temperature rise, and that there is a linear relation between the logarithm values of the diffusion coefficients and the reciprocal of the temperature. It is understood to be able to express the temperature dependency of the diffusion coefficient roughly by an Arrhenius equation as the velocity coefficient is provided as the diffusion coefficient. The specifications and characteristics of storage facilities under operation including dual purpose casks were investigated. Components subject to material degradation were examined. Based on literature survey, investigating a drop analysis incorporating with material degradation, the basic data to develop an analysis methodology was obtained

  16. Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, Kaushik [ORNL; Scaglione, John M [ORNL

    2015-01-01

    The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated keff margin varies from 0.05 to almost 0.3 Δkeff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

  17. Uraninite alteration in an oxidizing environment and its relevance to the disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Uraninite is a natural analogue for spent nuclear fuel because of similarities in structure and chemistry. Effective assessment of the long-term behavior of spent fuel in a geologic repository requires a knowledge of the corrosion products produced in that environment. Several important natural analogue sites are reviewed, illustrating a wide variety of environments from oxidizing to reducing, including, among others: Cigar Lake, Canada, a uraninite-bearing ore body at depth within a strictly reducing environment; the ore body has 'seen' extensive groundwater interaction with virtually no significant oxidation or mobilization of U apperent. Koongara, Australia is a highly altered uraninite-bearing ore body partially exposed to meteoric water; alteration at depth has resulted from interaction with groundwater having a somewhat reduced Eh compared to the surface. Uraninite, Pb-uranyl oxide hydrates and uranyl silicates control U solubility at depth; uranyl phosphates and U adsorption onto clays and FeMn-oxides control U solubility near the surface. Pocos de Caldas, Brazil displays a redox from moving through uraninite-bearing rocks near the surface and shows local remobilization of U. Oklo, Gabon, a uraninite- and coffinite-bearing ore body, locally affected by intense hydrothermal alteration during fission reactions, demonstrates restricted radionuclide and fission product transport within a reducing environment. A current study being conducted by the authors at Shinkolbwe, Zaire, a uraninite-bearing ore body exposed to highly oxidizing conditions at the surface, provides over 50 species of uranyl phases for detailed study, and illustrates a complex uranyl phase paragenesis over several million years, from earliest-formed uranyl oxide hydrates and uranyl silicates to later-formed uranyl phosphate. (au) (268 refs.)

  18. Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages

    International Nuclear Information System (INIS)

    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  19. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1998-09-01

    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  20. Nuclear Waste Risk Perceptions and Attitudes in Siting a Final Repository for Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    some extent unexplained as well. The reasons for the development towards more positive attitudes cannot be determined with any certainty at this point. However, it should be mentioned that attitude to nuclear power per se became more positive during the studied period, both in the two municipalities and in the nation as a whole. Nuclear power attitude was strongly correlated with policy intention in the present data (0.67). Yet, adding nuclear power attitude to the analysis of covariance model mentioned above did not result in a substantially weaker effect of the time factor (or of gender). This result suggests that whatever factor accounts for the changes observed, beyond the ones measured and investigated here, probably are area specific and possibly connected with the SKB program of information - a question for future research to answer