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Sample records for fuel snf process

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

    OLGUIN, L.J.

    2000-01-01

    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

  2. Spent Nuclear Fuel (SNF) Process Validation Technical Support Plan

    SEXTON, R.A.

    2000-03-13

    The purpose of Process Validation is to confirm that nominal process operations are consistent with the expected process envelope. The Process Validation activities described in this document are not part of the safety basis, but are expected to demonstrate that the process operates well within the safety basis. Some adjustments to the process may be made as a result of information gathered in Process Validation.

  3. Spent Nuclear Fuel (SNF) Process Validation Technical Support Plan

    SEXTON, R.A.

    2000-01-01

    The purpose of Process Validation is to confirm that nominal process operations are consistent with the expected process envelope. The Process Validation activities described in this document are not part of the safety basis, but are expected to demonstrate that the process operates well within the safety basis. Some adjustments to the process may be made as a result of information gathered in Process Validation

  4. 327 SNF fuel return to K-Basin quality process plan

    Ham, J.E.

    1998-01-01

    The B and W Hanford Company's (BWHC) 327 Facility, in the 300 Area of the Hanford Site, contains Spent Nuclear Fuel (SNF) single fuel element canisters (SFEC) and fuel remnant canisters (FRC) which are to be returned to K-Basin. Seven shipments of up to six fuel canisters will be loaded into the CNS 1-13G Cask and transported to 105-KE

  5. Spent Nuclear Fuel (SNF) Project Product Specification

    PAJUNEN, A.L.

    2000-01-01

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  6. Spent Nuclear Fuel (SNF) Project Product Specification

    PAJUNEN, A.L.

    2000-12-07

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  7. SNF Project Engineering Process Improvement Plan

    DESAI, S.P.

    2000-01-01

    This plan documents the SNF Project activities and plans to support its engineering process. It describes five SNF Project Engineering initiatives: new engineering procedures, qualification cards process; configuration management, engineering self assessments, and integrated schedule for engineering activities

  8. Spent Nuclear Fuel (SNF) Project Execution Plan

    LEROY, P.G.

    2000-01-01

    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

  9. Spent Nuclear Fuel (SNF) Project Execution Plan

    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.

  10. Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB) Process Flow Diagram Mass Balance Calculations

    KLEM, M.J.

    2000-01-01

    The purpose of these calculations is to develop the material balances for documentation of the Canister Storage Building (CSB) Process Flow Diagram (PFD) and future reference. The attached mass balances were prepared to support revision two of the PFD for the CSB. The calculations refer to diagram H-2-825869

  11. Spent Nuclear Fuel (SNF) Project Multi Canister Overpack (MCO) Process Flow Diagram Mass Balance Calculations

    KLEM, M.J.

    2000-01-01

    The purpose of this calculation document is to develop the bases for the material balances of the Multi-Canister Overpack (MCO) Level 1 Process Flow Diagram (PFD). The attached mass balances support revision two of the PFD for the MCO and provide future reference

  12. EFFICACY OF FILTRATION PROCESSES TO OBTAIN WATER CLARITY AT K EAST SPENT NUCLEAR FUEL (SNF) BASIN

    DUNCAN JB

    2006-09-28

    The objective is to provide water clarity to the K East Basin via filtration processes. Several activities are planned that will challenge not only the capacity of the existing ion exchange modules to perform as needed but also the current filtration system to maintain water clarity. Among the planned activities are containerization of sludge, removal of debris, and hydrolasing the basin walls to remove contamination.

  13. EFFICACY OF FILTRATION PROCESSES TO OBTAIN WATER CLARITY AT K EAST SPENT NUCLEAR FUEL (SNF) BASIN

    DUNCAN JB

    2006-01-01

    The objective is to provide water clarity to the K East Basin via filtration processes. Several activities are planned that will challenge not only the capacity of the existing ion exchange modules to perform as needed but also the current filtration system to maintain water clarity. Among the planned activities are containerization of sludge, removal of debris, and hydrolasing the basin walls to remove contamination

  14. Human error prediction and countermeasures based on CREAM in spent nuclear fuel (SNF) transportation

    Kim, Jae San

    2007-02-01

    Since the 1980s, in order to secure the storage capacity of spent nuclear fuel (SNF) at NPPs, SNF assemblies have been transported on-site from one unit to another unit nearby. However in the future the amount of the spent fuel will approach capacity in the areas used, and some of these SNFs will have to be transported to an off-site spent fuel repository. Most SNF materials used at NPPs will be transported by general cargo ships from abroad, and these SNFs will be stored in an interim storage facility. In the process of transporting SNF, human interactions will involve inspecting and preparing the cask and spent fuel, loading the cask onto the vehicle or ship, transferring the cask as well as storage or monitoring the cask. The transportation of SNF involves a number of activities that depend on reliable human performance. In the case of the transport of a cask, human errors may include spent fuel bundle misidentification or cask transport accidents among others. Reviews of accident events when transporting the Radioactive Material (RAM) throughout the world indicate that human error is the major causes for more than 65% of significant events. For the safety of SNF transportation, it is very important to predict human error and to deduce a method that minimizes the human error. This study examines the human factor effects on the safety of transporting spent nuclear fuel (SNF). It predicts and identifies the possible human errors in the SNF transport process (loading, transfer and storage of the SNF). After evaluating the human error mode in each transport process, countermeasures to minimize the human error are deduced. The human errors in SNF transportation were analyzed using Hollnagel's Cognitive Reliability and Error Analysis Method (CREAM). After determining the important factors for each process, countermeasures to minimize human error are provided in three parts: System design, Operational environment, and Human ability

  15. SNF project engineering process improvement plan

    DESAI, S.P.

    1999-01-01

    This Engineering Process Improvement Plan documents the activities and plans to be taken by the SNF Project to support its engineering process and to produce a consolidated set of engineering procedures that are fully compliant with the requirements of HNF-PRO-1819. All new procedures will be issued and implemented by September 30, 1999

  16. Spent Nuclear Fuel (SNF) Removal Campaign Plan

    PAJUNEN, A.L.

    2000-01-01

    The overall operation of the Spent Nuclear Fuel Project will include fuel removal, sludge removal, debris removal, and deactivation transition activities. Figure 1-1 provides an overview of the current baseline operating schedule for project sub-systems, indicating that a majority of fuel removal activities are performed over an approximately three-and-one-half year time period. The purpose of this document is to describe the strategy for operating the fuel removal process systems. The campaign plan scope includes: (1) identifying a fuel selection sequence during fuel removal activities, (2) identifying MCOs that are subjected to extra testing (process validation) and monitoring, and (3) discussion of initial MCO loading and monitoring in the Canister Storage Building (CSB). The campaign plan is intended to integrate fuel selection requirements for handling special groups of fuel within the basin (e.g., single pass reactor fuel), process validation activities identified for process systems, and monitoring activities during storage

  17. SNF project engineering process improvement plan

    KELMENSON, R.L.

    1999-01-01

    This Engineering Process Improvement Plan documents the activities and plans to be taken by the SNF Project (the Project) to support its engineering process and to produce a consolidated set of engineering procedures that are fully compliant with the requirements of HNF-PRO-1819 (1819). These requirements are imposed on all engineering activities performed for the Project and apply to all life-cycle stages of the Project's systems, structures and components (SSCs). This Plan describes the steps that will be taken by the Project during the transition period to ensure that new procedures are effectively integrated into the Project's work process as these procedures are issued. The consolidated procedures will be issued and implemented by September 30, 1999

  18. SNF fuel retrieval sub project safety analysis document

    BERGMANN, D.W.

    1999-01-01

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed

  19. SNF fuel retrieval sub project safety analysis document

    BERGMANN, D.W.

    1999-02-24

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed.

  20. Preparation for the Recovery of Spent Nuclear Fuel (SNF) at Andreeva Bay, North West Russia - 13309

    Field, D.; McAtamney, N.

    2013-01-01

    Andreeva Bay is located near Murmansk in the Russian Federation close to the Norwegian border. The ex-naval site was used to de-fuel nuclear-powered submarines and icebreakers during the Cold War. Approximately 22,000 fuel assemblies remain in three Dry Storage Units (DSUs) which means that Andreeva Bay has one of the largest stockpiles of highly enriched spent nuclear fuel (SNF) in the world. The high contamination and deteriorating condition of the SNF canisters has made improvements to the management of the SNF a high priority for the international community for safety, security and environmental reasons. International Donors have, since 2002, provided support to projects at Andreeva concerned with improving the management of the SNF. This long-term programme of work has been coordinated between the International Donors and responsible bodies within the Russian Federation. Options for the safe and secure management of SNF at Andreeva Bay were considered in 2004 and developed by a number of Russian Institutes with international participation. This consisted of site investigations, surveys and studies to understand the technical challenges. A principal agreement was reached that the SNF would be removed from the site altogether and transported to Russia's reprocessing facility at Mayak in the Urals. The analytical studies provided the information necessary to develop the construction plan for the site. Following design and regulatory processes, stakeholders endorsed the technical solution in April 2007. This detailed the processes, facilities and equipment required to safely remove the SNF and identified other site services and support facilities required on the site. Implementation of this strategy is now well underway with the facilities in various states of construction. Physical works have been performed to address the most urgent tasks including weather protection over one of the DSUs, installation of shielding over the cells, provision of radiation

  1. Preparation for the Recovery of Spent Nuclear Fuel (SNF) at Andreeva Bay, North West Russia - 13309

    Field, D.; McAtamney, N. [Nuvia Limited (United Kingdom)

    2013-07-01

    Andreeva Bay is located near Murmansk in the Russian Federation close to the Norwegian border. The ex-naval site was used to de-fuel nuclear-powered submarines and icebreakers during the Cold War. Approximately 22,000 fuel assemblies remain in three Dry Storage Units (DSUs) which means that Andreeva Bay has one of the largest stockpiles of highly enriched spent nuclear fuel (SNF) in the world. The high contamination and deteriorating condition of the SNF canisters has made improvements to the management of the SNF a high priority for the international community for safety, security and environmental reasons. International Donors have, since 2002, provided support to projects at Andreeva concerned with improving the management of the SNF. This long-term programme of work has been coordinated between the International Donors and responsible bodies within the Russian Federation. Options for the safe and secure management of SNF at Andreeva Bay were considered in 2004 and developed by a number of Russian Institutes with international participation. This consisted of site investigations, surveys and studies to understand the technical challenges. A principal agreement was reached that the SNF would be removed from the site altogether and transported to Russia's reprocessing facility at Mayak in the Urals. The analytical studies provided the information necessary to develop the construction plan for the site. Following design and regulatory processes, stakeholders endorsed the technical solution in April 2007. This detailed the processes, facilities and equipment required to safely remove the SNF and identified other site services and support facilities required on the site. Implementation of this strategy is now well underway with the facilities in various states of construction. Physical works have been performed to address the most urgent tasks including weather protection over one of the DSUs, installation of shielding over the cells, provision of radiation

  2. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    PAJUNEN, A.L.

    2000-09-20

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project.

  3. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    PAJUNEN, A.L.

    2000-01-01

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project

  4. Spent Nuclear Fuel (SNF) Project Safety Basis Implementation Strategy

    TRAWINSKI, B.J.

    2000-01-01

    The objective of the Safety Basis Implementation is to ensure that implementation of activities is accomplished in order to support readiness to move spent fuel from K West Basin. Activities may be performed directly by the Safety Basis Implementation Team or they may be performed by other organizations and tracked by the Team. This strategy will focus on five key elements, (1) Administration of Safety Basis Implementation (general items), (2) Implementing documents, (3) Implementing equipment (including verification of operability), (4) Training, (5) SNF Project Technical Requirements (STRS) database system

  5. Human Error Prediction and Countermeasures based on CREAM in Loading and Storage Phase of Spent Nuclear Fuel (SNF)

    Kim, Jae San; Kim, Min Su; Jo, Seong Youn

    2007-01-01

    With the steady demands for nuclear power energy in Korea, the amount of accumulated SNF has inevitably increased year by year. Thus far, SNF has been on-site transported from one unit to a nearby unit or an on-site dry storage facility. In the near future, as the amount of SNF generated approaches the capacity of these facilities, a percentage of it will be transported to another SNF storage facility. In the process of transporting SNF, human interactions involve inspecting and preparing the cask and spent fuel, loading the cask, transferring the cask and storage or monitoring the cask, etc. So, human actions play a significant role in SNF transportation. In analyzing incidents that have occurred during transport operations, several recent studies have indicated that 'human error' is a primary cause. Therefore, the objectives of this study are to predict and identify possible human errors during the loading and storage of SNF. Furthermore, after evaluating human error for each process, countermeasures to minimize human error are deduced

  6. A contingency safe, responsible, economic, increased capacity spent nuclear fuel (SNF) advance fuel cycle

    Levy, S.

    2008-01-01

    The purpose of this paper is to have an Advanced Light Water (LWR) fuel cycle and an associated development program to provide a contingency plan to the current DOE effort to license once-through spent Light Water Reactor (LWR) fuel for disposition at Yucca Mountain (YM). The intent is to fully support the forthcoming June 2008 DOE submittal to the Nuclear Regulatory Commission (NRC) based upon the latest DOE draft DOE/EIS-0250F-SID dated October 2007 which shows that the latest DOE YM doses would readily satisfy the anticipated NRC and Environmental Protection Agency (EP) standards. The proposed Advance Fuel Cycle can offer potential resolution of obstacles that might arise during the NRC review and, particularly, during the final hearings process to be held in Nevada. Another reason for the proposed concept is that a substantial capacity growth of the YM repository will be necessary to accommodate the SNF of Advance Light Water Reactors (ALWRs) currently under consideration for United States (U.S.) electricity production (1) and the results of the recently issued study by the Electric Power Research Institute (EPRI) to reduce CO 2 emissions (2). That study predicts that by 2030 U.S. nuclear power generation would grow by 64 Gigawatt electrical (GWe) and account for 25.5 percent of the overall U.S. electrical generation. The current annual SNF once-through fuel cycle accumulation would rise from 2000-2100 MT (Metric Tons) to about 3480 MT in 2030 and the total SNF inventory, would reach nearly 500,000 MT by 2100 if U. S. nuclear power continues to grow at 1.1 percent per year after 2030. That last projection does not account for any SNF reduction due to increased fuel burnup or any increased capacity needed 'to establish supply Global Nuclear Energy Partnership (GNEP,) arrangements among nations to provide nuclear fuel and taking back spent fuel for recycling without spreading enrichment and reprocessing technologies' (3). The anticipated capacity of 120 MT

  7. Plan for Characterization of K Basin Spent Nuclear Fuel (SNF) and Sludge (OCRWM)

    TRIMBLE, D.J.

    2000-01-01

    This is an update of the plan for the characterization of spent nuclear fuel (SNF) and sludge stored in the Hanford K West and K East Basins. The purpose of the characterization program is to provide fuel and sludge data in support of the SNF Project in the effort to remove the fuel from the K Basins and place it into dry storage. Characterization of the K Basin fuel and sludge was initiated in 1994 and has been guided by the characterization plans (Abrefah 1994, Lawrence 1995a, Lawrence 1995b) and the characterization program management plan (PMP) (Lawrence 1995c, Lawrence 1998, Trimble 1999). The fuel characterization was completed in 1999. Summaries of these activities were documented by Lawrence (1999) and Suyama (1999). Lawrence (1999) is a summary report providing a road map to the detailed documentation of the fuel characterization. Suyama (1999) provides a basis for the limited characterization sample size as it relates to supporting design limits and the operational safety envelope for the SNF Project. The continuing sludge characterization is guided by a data quality objective (DQO) (Makenas 2000) and a sampling and analysis plan (SAP) (Baker, Welsh and Makenas 2000) The original intent of the characterization program was ''to provide bounding behavior for the fuel'' (Lawrence 1995a). To accomplish this objective, a fuel characterization program was planned that would provide data to augment data from the literature. The program included in-situ examinations of the stored fuel and laboratory testing of individual elements and small samples of fuel (Lawrence 1995a). Some of the planned tests were scaled down or canceled due to the changing needs of the SNF Project. The fundamental technical basis for the process that will be used to place the K Basin fuel into dry storage was established by several key calculations. These calculations characterized nominal and bounding behavior of fuel in Multi-Canister Overpacks (MCOs) during processing and storage

  8. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    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.

  9. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    IRWIN, J.J.

    2000-01-01

    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

  10. Problems and solutions of the spent nuclear fuel (SNF) at Kozloduy NPP

    Jordanov, J.

    2003-01-01

    There are two options concerning spent nuclear fuel: to return it back to Russia for reprocessing or to store it on the site until we decide what to do with it. In both options prior to the shutting down of each reactor the Spent Fuel Pool thereto should be vacated (the filling in of the equipment at present is illustrated) and the Spent Fuel Storage Facility (SFSF) should also be vacated after the stop of the last nuclear facility on the site in order to be reequipped for permanent storage of the highly active wastes which will be returned in the country, if we submit the fuel for reprocessing; or of SNF, if we decide to leave them ultimately in Bulgaria. The difference is mainly in the quantities which will permanently remain here, respectively the volumes required for their storage and the funds necessary for the implementation of the processes. The pool volumes filling in both variants is also illustrated and the SFSF will be filled by 2008, if no fuel is transported.Costs of the SNF transport to Russia and investment costs of dry storage of SNF from pools 1 - 4 are present. The costs are visibly lower compared to those in the case of return of the fuel. However, these are only investments for construction and equipment of the buildings and storage containers. The costs related to their servicing are not included, and it should be taken into account that in approximately 50 years we will have to seek solution for their permanent storage. Despite the material costs to be incurred now for the implementation of the option with the return of the fuel, this is the more worthy way to resolve the problem. In accordance with the ethic principles in the nuclear energy, the burdens arising as a result of the use of nuclear facilities should be covered by the generation consuming the benefits from it

  11. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    IRWIN, J.J.

    2000-01-01

    The mission of the Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying Facility (CVDF) is to achieve the earliest possible removal of free water from Multi-Canister Overpacks (MCOs). The MCOs contain metallic uranium SNF that have been removed from the 100K Area fuel storage water basins (i.e., the K East and K West Basins) at the US. Department of Energy Hanford Site in Southeastern Washington state. Removal of free water is necessary to halt water-induced corrosion of exposed uranium surfaces and to allow the MCOs and their SNF payloads to be safely transported to the Hanford Site 200 East Area and stored within the SNF Project Canister Storage Building (CSB). The CVDF is located within a few hundred yards of the basins, southwest of the 165KW Power Control Building and the 105KW Reactor Building. The site area required for the facility and vehicle circulation is approximately 2 acres. Access and egress is provided by the main entrance to the 100K inner area using existing roadways. The CVDF will remove free. water from the MCOs to reduce the potential for continued fuel-water corrosion reactions. The cold vacuum drying process involves the draining of bulk water from the MCO and subsequent vacuum drying. The MCO will be evacuated to a pressure of 8 torr or less and backfilled with an inert gas (helium). The MCO will be sealed, leak tested, and then transported to the CSB within a sealed shipping cask. (The MCO remains within the same shipping Cask from the time it enters the basin to receive its SNF payload until it is removed from the Cask by the CSB MCO handling machine.) The CVDF subproject acquired the required process systems, supporting equipment, and facilities. The cold vacuum drying operations result in an MCO containing dried fuel that is prepared for shipment to the CSB by the Cask transportation system. The CVDF subproject also provides equipment to dispose of solid wastes generated by the cold vacuum drying process and transfer process water removed

  12. K Basins Spent Nuclear Fuel (SNF) Project Safety Analysis Report for Packaging (SARP) approval plan

    1995-01-01

    This document delineates the plan for preparation, review, and approval of the K Basins Spent Nuclear Fuel (SNF) Packaging Design Criteria (PDC) document and the on-site Safety Analysis Report for Packaging (SARP). The packaging addressed in these documents is used to transport SNF in a Multi- canister Overpack (MCO) configuration

  13. Interrelation of technologies for RW preparation and sites for final isolation of the wastes from pyrochemical processing of SNF

    Gupalo, V.S.; Chistyakov, V.N. [JSC - Design-Prospecting and Scientific-Research Institute of Industrial Technology -, Kashirskoye Highway, 33, Moscow 115409 (Russian Federation); Kormilitsyn, M.V.; Kormilitsyna, L.A. [JSC - State Scientific Center - Research Institute of Atomic Reactors -, Ulyanovsk region, Dimitrovgrad - 10, 433510 (Russian Federation)

    2013-07-01

    For the justification of engineering solutions and practical testing of the radiochemical component of the perspective nuclear power complex with on-site variant of nuclear fuel cycle (NFC), it is planned to establish a multi-functional research-development complex (MFCRC) for radiochemical processing of spent nuclear fuels (SNF) from fast reactors. MFCRC is being established at the NIIAR site, it comprises technological process lines, where innovation pyro-electrochemical and hydrometallurgical technologies are realized, with an option for closing the inter-chain material flows for testing the combined radiochemically converted materials. The technological flowchart for processing at the MFCRC is subdivided into 3 segments: -) complex of the lead operations for dismantling the fuel elements (FE) and fuel assemblies (FA), -) pyrochemical extraction flowchart for processing SNF, and -) hydrometallurgical flowchart for processing SNF. The engineered solutions for the management and disposition of the radioactive wastes from MFCRC are reviewed.

  14. Volumes, Masses, and Surface Areas for Shippingport LWBR Spent Nuclear Fuel in a DOE SNF Canister

    J.W. Davis

    1999-01-01

    The purpose of this calculation is to estimate volumes, masses, and surface areas associated with (a) an empty Department of Energy (DOE) 18-inch diameter, 15-ft long spent nuclear fuel (SNF) canister, (b) an empty DOE 24-inch diameter, 15-ft long SNF canister, (c) Shippingport Light Water Breeder Reactor (LWBR) SNF, and (d) the internal basket structure for the 18-in. canister that has been designed specifically to accommodate Seed fuel from the Shippingport LWBR. Estimates of volumes, masses, and surface areas are needed as input to structural, thermal, geochemical, nuclear criticality, and radiation shielding calculations to ensure the viability of the proposed disposal configuration

  15. Spent Nuclear Fuel (SNF) Startup Plan to Operations

    GREGORY, J.R.

    2000-01-01

    This plan defines the approach that will be used to ensure the transition from initial startup to normal operations of the SNF operations--are performed in a safe, controlled, and deliberate manner. It provides a phased approach that bridges the operations between the completion of the ORR and the return to normal operations. This plan includes management oversight and administrative controls to be implemented and then reduced in a controlled manner until normal operations are authorized by SNF Management

  16. Spent Nuclear Fuel (SNF) Project Design Basis Capacity Study

    CLEVELAND, K.J.

    2000-01-01

    This study of the design basis capacity of process systems was prepared by Fluor Federal Services for the Spent Nuclear Fuel Project. The evaluation uses a summary level model of major process sub-systems to determine the impact of sub-system interactions on the overall time to complete fuel removal operations. The process system model configuration and time cycle estimates developed in the original version of this report have been updated as operating scenario assumptions evolve. The initial document released in Fiscal Year (FY) 1996 varied the number of parallel systems and transport systems over a wide range, estimating a conservative design basis for completing fuel processing in a two year time period. Configurations modeling planned operations were updated in FY 1998 and FY 1999. The FY 1998 Base Case continued to indicate that fuel removal activities at the basins could be completed in slightly over 2 years. Evaluations completed in FY 1999 were based on schedule modifications that delayed the start of KE Basin fuel removal, with respect to the start of KW Basin fuel removal activities, by 12 months. This delay resulted in extending the time to complete all fuel removal activities by 12 months. However, the results indicated that the number of Cold Vacuum Drying (CVD) stations could be reduced from four to three without impacting the projected time to complete fuel removal activities. This update of the design basis capacity evaluation, performed for FY 2000, evaluates a fuel removal scenario that delays the start of KE Basin activities such that staffing peaks are minimized. The number of CVD stations included in all cases for the FY 2000 evaluation is reduced from three to two, since the scenario schedule results in minimal time periods of simultaneous fuel removal from both basins. The FY 2000 evaluation also considers removal of Shippingport fuel from T Plant storage and transfer to the Canister Storage Building for storage

  17. Spent Nuclear Fuel (SNF) Bounding Drop Support Calculations

    CHENAULT, D.M.

    1999-01-01

    This report evaluates different drop heights, concrete and other impact media to which the transport package and/or the MCO is dropped. A prediction method is derived for estimating the resultant impact factor for determining the bounding drop case for the SNF Project

  18. Characterization Program Management Plan for Hanford K Basin Spent Nuclear Fuel (SNF) (OCRWM)

    BAKER, R.B.; TRIMBLE, D.J.

    2000-01-01

    The management plan developed to characterize the K Basin spent nuclear fuel (SNF) and sludge was originally developed for Westinghouse Hanford Company and Pacific Northwest National Laboratory to work together on a program to provide characterization data to support removal, conditioning, and subsequent dry storage of the SNF stored at the Hanford K Basins. The plan also addressed necessary characterization for the removal, transport, and storage of the sludge from the Hanford K Basins. This plan was revised in 1999 (i.e., Revision 2) to incorporate actions necessary to respond to the deficiencies revealed as the result of Quality Assurance surveillances and audits in 1999 with respect to the fuel characterization activities. Revision 3 to this Program Management Plan responds to a Worker Assessment resolution determined in Fical Year 2000. This revision includes an update to current organizational structures and other revisions needed to keep this management plan consistent with the current project scope. The plan continues to address both the SNF and the sludge accumulated at K Basins. Most activities for the characterization of the SNF have been completed. Data validation, Office of Civilian Radioactive Waste Management (OCRWM) document reviews, and OCRWM data qualification are the remaining SNF characterization activities. The transport and storage of K Basin sludge are affected by recent path forward revisions. These revisions require additional laboratory analyses of the sludge to complete the acquisition of required supporting engineering data. Hence, this revision of the management plan provides the overall work control for these remaining SNF and sludge characterization activities given the current organizational structure of the SNF Project

  19. Spent Nuclear Fuel (SNF) Cold Vacuum Drying (CVD) Facility Operations Manual; FINAL

    IRWIN, J.J.

    1999-01-01

    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

  20. Spent Nuclear Fuel (SNF) Cold Vacuum Drying (CVD) Facility Operations Manual

    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.

  1. Methodology of fuel cycles long-term safety assessment of SNF/HLW geological disposal

    Pritrsky, J.

    2008-02-01

    Methodology for the long-term safety assessment of nuclear fuel cycles is given in the presented doctoral thesis. The aim of work was to develop a geological repository model for disposal of spent nuclear fuel (SNF) and high level waste (HLW) using an appropriate software code able to calculate the influence of partitioning and transmutation in advanced fuel cycles. The first step in this process was specifying of indicators which can be used to quantify the radiological impact of each fuel cycle. Indicators such as annual effective dose and radiotoxicity of inventory have been quantitatively analysed to determine the potential risk and radiological consequences associated with production of SNF/HLW. Advanced fuel types bring a number of advantages in comparison to uranium oxide fuel UO 2 used worldwide nowadays in terms of safety improvement due to minor actinides transmutation and non-proliferation aspects as well. Within the scope of work, three different fuel cycles are compared from the point of view of long-term safety of deep geological repository. The first considered fuel cycle is the currently used open fuel cycle (UOX) which uses only U-FA (Uranium Fuel Assembly). The second assessed cycle is a closed fuel cycle (MOX) with MOX-FA (Mixed OXides Fuel Assembly) and the third considered one is a partially closed fuel cycle (IMF) with IMC-FA (Inert Matrix Combined Fuel Assembly). Description and input data of advanced fuel cycles have been gained by participation in the EC project RED-IMPACT. Results were calculated using code AMBER, which is a flexible software tool that allows building dynamic compartmental models to represent the migration and fate of contaminants in a system, for example in the surface and sub-surface environment. Contaminants in solid, liquid and gaseous phases can be considered. AMBER gives the user the flexibility to define any number of compartments; any number of contaminants and associated decays; deterministic, probabilistic and

  2. Methodology of fuel cycles long-term safety assessment of SNF/HLW geological disposal

    Pritrsky, J.

    2008-01-01

    Methodology for the long-term safety assessment of nuclear fuel cycles is given in the presented doctoral thesis. The aim of work was to develop a geological repository model for disposal of spent nuclear fuel (SNF) and high level waste (HLW) using an appropriate software code able to calculate the influence of partitioning and transmutation in advanced fuel cycles. The first step in this process was specifying of indicators which can be used to quantify the radiological impact of each fuel cycle. Indicators such as annual effective dose and radiotoxicity of inventory have been quantitatively analysed to determine the potential risk and radiological consequences associated with production of SNF/HLW. Advanced fuel types bring a number of advantages in comparison to uranium oxide fuel UO 2 used worldwide nowadays in terms of safety improvement due to minor actinides transmutation and non-proliferation aspects as well. Within the scope of work, three different fuel cycles are compared from the point of view of long-term safety of deep geological repository. The first considered fuel cycle is the currently used open fuel cycle (UOX) which uses only U-FA (Uranium Fuel Assembly). The second assessed cycle is a closed fuel cycle (MOX) with MOX-FA (Mixed OXides Fuel Assembly) and the third considered one is a partially closed fuel cycle (IMF) with IMC-FA (Inert Matrix Combined Fuel Assembly). Description and input data of advanced fuel cycles have been gained by participation in the EC project RED-IMPACT. Results were calculated using code AMBER, which is a flexible software tool that allows building dynamic compartmental models to represent the migration and fate of contaminants in a system, for example in the surface and sub-surface environment. Contaminants in solid, liquid and gaseous phases can be considered. AMBER gives the user the flexibility to define any number of compartments; any number of contaminants and associated decays; deterministic, probabilistic and

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

    SCHULTZ, M.V.

    2000-01-01

    As part of the Spent Nuclear Fuel (SNF) storage basin clean-up project, sludge that has accumulated in the K Basins due to corrosion of damaged irradiated N Reactor will be loaded into containers and placed in interim storage. The Hanford Site Treatment Complex (T Plant) has been identified as the location where the sludge will be stored until final disposition of the material occurs. Long term storage of sludge from the K Basin fuel storage facilities requires identification and analysis of potential accidents involving sludge storage in T Plant. This report is prepared as the initial step in the safety assurance process described in DOE Order 5480.23, Nuclear Safety Analysis Reports and HNF-PRO-704, Hazards and Accident Analysis Process. This report documents the evaluation of potential hazards and off-normal events associated with sludge storage activities. This information will be used in subsequent safety analyses, design, and operations procedure development to ensure safe storage. The hazards evaluation for the storage of SNF sludge in T-Plant used the Hazards and Operability Analysis (HazOp) method. The hazard evaluation identified 42 potential hazardous conditions. No hazardous conditions involving hazardous/toxic chemical concerns were identified. Of the 42 items identified in the HazOp study, eight were determined to have potential for onsite worker consequences. No items with potential offsite consequences were identified in the HazOp study. Hazardous conditions with potential onsite worker or offsite consequences are candidates for quantitative consequence analysis. The hazardous conditions with potential onsite worker consequences were grouped into two event categories, Container failure due to overpressure - internal to T Plant, and Spill of multiple containers. The two event categories will be developed into accident scenarios that will be quantitatively analyzed to determine release consequences. A third category, Container failure due to

  4. White Paper: Multi-purpose canister (MPC) for DOE-owned spent nuclear fuel (SNF)

    Knecht, D.A.

    1994-04-01

    The paper examines the issue, What are the advantages, disadvantages, and other considerations for using the MPC concept as part of the strategy for interim storage and disposal of DOE-owned SNF? The paper is based in part on the results of an evaluation made for the DOE National Spent Fuel Program by the Waste Form Barrier/Canister Team, which is composed of knowledgeable DOE and DOE-contractor personnel. The paper reviews the MPC and DOE SNF status, provides criteria and other considerations applicable to the issue, and presents an evaluation, conclusions, and recommendations. The primary conclusion is that while most of DOE SNF is not currently sufficiently characterized to be sealed into an MPC, the advantages of standardized packages in handling, reduced radiation exposure, and improved human factors should be considered in DOE SNF program planning. While the design of MPCs for DOE SNF are likely premature at this time, the use of canisters should be considered which are consistent with interim storage options and the MPC design envelope

  5. National spent fuel program preliminary report RCRA characteristics of DOE-owned spent nuclear fuel DOE-SNF-REP-002. Revision 3

    1995-07-01

    This report presents information on the preliminary process knowledge to be used in characterizing all Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) types that potentially exhibit a Resource Conservation and Recovery Act (RCRA) characteristic. This report also includes the process knowledge, analyses, and rationale used to preliminarily exclude certain SNF types from RCRA regulation under 40 CFR section 261.4(a)(4), ''Identification and Listing of Hazardous Waste,'' as special nuclear and byproduct material. The evaluations and analyses detailed herein have been undertaken as a proactive approach. In the event that DOE-owned SNF is determined to be a RCRA solid waste, this report provides general direction for each site regarding further characterization efforts. The intent of this report is also to define the path forward to be taken for further evaluation of specific SNF types and a recommended position to be negotiated and established with regional and state regulators throughout the DOE Complex regarding the RCRA-related policy issues

  6. National spent fuel program preliminary report RCRA characteristics of DOE-owned spent nuclear fuel DOE-SNF-REP-002. Revision 3

    NONE

    1995-07-01

    This report presents information on the preliminary process knowledge to be used in characterizing all Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) types that potentially exhibit a Resource Conservation and Recovery Act (RCRA) characteristic. This report also includes the process knowledge, analyses, and rationale used to preliminarily exclude certain SNF types from RCRA regulation under 40 CFR {section}261.4(a)(4), ``Identification and Listing of Hazardous Waste,`` as special nuclear and byproduct material. The evaluations and analyses detailed herein have been undertaken as a proactive approach. In the event that DOE-owned SNF is determined to be a RCRA solid waste, this report provides general direction for each site regarding further characterization efforts. The intent of this report is also to define the path forward to be taken for further evaluation of specific SNF types and a recommended position to be negotiated and established with regional and state regulators throughout the DOE Complex regarding the RCRA-related policy issues.

  7. Radionuclide Inventories for DOE SNF Waste Stream and Uranium/Thorium Carbide Fuels

    K.L. Goluoglu

    2000-01-01

    The objective of this calculation is to generate radionuclide inventories for the Department of Energy (DOE) spent nuclear fuel (SNF) waste stream destined for disposal at the potential repository at Yucca Mountain. The scope of this calculation is limited to the calculation of two radionuclide inventories; one for all uranium/thorium carbide fuels in the waste stream and one for the entire waste stream. These inventories will provide input in future screening calculations to be performed by Performance Assessment to determine important radionuclides

  8. Technical Basis - Spent Nuclear Fuels (SNF) Project Radiation and Contamination Trending Program

    ELGIN, J.C.

    2000-01-01

    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

  9. Technology development for DOE SNF management

    Hale, D.L.; Einziger, R.E.; Murphy, J.R.

    1995-01-01

    This paper describes the process used to identify technology development needs for the same management of spent nuclear fuel (SNF) in the US Department of Energy (DOE) inventory. Needs were assessed for each of the over 250 fuel types stores at DOE sites around the country for each stage of SNF management--existing storage, transportation, interim storage, and disposal. The needs were then placed into functional groupings to facilitate integration and collaboration among the sites

  10. DESIGN VERIFICATION REPORT SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB)

    BAZINET, G.D.

    2003-02-12

    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. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Revision 3 of this document incorporates MCO Cover Cap Assembly welding verification activities. Verification activities for the installed and operational SSCs have been completed.

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

    BAZINET, G.D.

    2001-05-15

    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. Revision 1 documented 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 requirements are noted in section 3

  12. DESIGN VERIFICATION REPORT SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB)

    BAZINET, G.D.

    2003-01-01

    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. Revision 1 documented verification actions that were pending at the time the initial report was prepared. Revision 3 of this document incorporates MCO Cover Cap Assembly welding verification activities. Verification activities for the installed and operational SSCs have been completed

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

    BAZINET, G.D.

    2001-01-01

    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. Revision 1 documented 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 requirements are noted in section 3.1.5 and will be

  14. Safety assessment for a potential SNF repository and its implication to the proliferation resistance nuclear fuel cycle

    Hwang, Y.; Jeong, M.S.; Seo, C.S.

    2007-01-01

    KAERI is developing the pyro-process technology to minimize the burden on permanent disposal of spent nuclear fuel. In addition, KAERI has developed the Korean Reference System for potential spent nuclear fuel disposal since 1997. The deep geologic disposal system is composed of a multi-barrier system in a crystalline rock to dispose of 36,000 MT of spent nuclear fuel (SNF) from a CANDU and a PWR. Quite recently, introduction of advanced nuclear fuel cycles such as pyro-processing is a big issue to solve the everlasting disposal problem and to assure the sustainable supply of fuel for reactors. To compare the effect of direct disposal of SNF with that of the high level waste disposal for waste generated from the advanced nuclear fuel cycles, the total system performance assessment for two different schemes is developed; one for direct disposal of SNF and the other for the introduction of the pyro-processing and direct disposal CANDU spent nuclear fuel. The safety indicators to assess the environmental friendliness of the disposal option are annual individual doses, toxicities and risks. Even though many scientists use the toxicity to understand the environmental friendliness of the disposal, scientifically the annual individual doses or risks are meaningful indicators for it. The major mechanisms to determine the doses and risks for direct disposal are as follows: (1) Dissolution mechanisms of uranium dioxides which control the dissolution of most nuclides such as TRU's and most parts of fission products. (2) Instant release fraction of highly soluble nuclides such as I-129, C-135, Tc-99, and others. (3) Retardation and dilution effect of natural and engineered barriers. (4) Dilution effect in the biosphere. The dominant nuclide is I-129 which follows both congruent and instantaneous release modes. Since its long half life associated with the instantaneous release I-129 is dominant well beyond one million. The impact of the TRU's is negligible until the significant

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

    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

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

    BAZINET, G.D.

    2000-01-01

    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 requirements are noted

  17. Fuel processing

    Allardice, R.H.

    1990-01-01

    The technical and economic viability of the fast breeder reactor as an electricity generating system depends not only upon the reactor performance but also on a capability to recycle plutonium efficiently, reliably and economically through the reactor and fuel cycle facilities. Thus the fuel cycle is an integral and essential part of the system. Fuel cycle research and development has focused on demonstrating that the challenging technical requirements of processing plutonium fuel could be met and that the sometimes conflicting requirements of the fuel developer, fuel fabricator and fuel reprocessor could be reconciled. Pilot plant operation and development and design studies have established both the technical and economic feasibility of the fuel cycle but scope for further improvement exists through process intensification and flowsheet optimization. These objectives and the increasing processing demands made by the continuing improvement to fuel design and irradiation performance provide an incentive for continuing fuel cycle development work. (author)

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

    PICKETT, W.W.

    2000-09-22

    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. Because this sub-project is still in the construction/start-up phase, all verification activities have not yet been performed (e.g., canister cover cap and welding fixture system verification, MCO Internal Gas Sampling equipment verification, and As-built verification.). The verification activities identified in this report that still are to be performed will be added to the start-up punchlist and tracked to closure.

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

    PICKETT, W.W.

    2000-01-01

    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. Because this sub-project is still in the construction/start-up phase, all verification activities have not yet been performed (e.g., canister cover cap and welding fixture system verification, MCO Internal Gas Sampling equipment verification, and As-built verification.). The verification activities identified in this report that still are to be performed will be added to the start-up punchlist and tracked to closure

  20. DOE SNF technology development necessary for final disposal

    Hale, D.L.; Fillmore, D.L.; Windes, W.E.

    1996-01-01

    Existing technology is inadequate to allow safe disposal of the entire inventory of US Department of Energy (DOE) spent nuclear fuel (SNF). Needs for SNF technology development were identified for each individual fuel type in the diverse inventory of SNF generated by past, current, and future DOE materials production, as well as SNF returned from domestic and foreign research reactors. This inventory consists of 259 fuel types with different matrices, cladding materials, meat composition, actinide content, and burnup. Management options for disposal of SNF include direct repository disposal, possible including some physical or chemical preparation, or processing to produce a qualified waste form by using existing aqueous processes or new treatment processes. Technology development needed for direct disposal includes drying, mitigating radionuclide release, canning, stabilization, and characterization technologies. While existing aqueous processing technology is fairly mature, technology development may be needed to apply one of these processes to SNF different than for which the process was originally developed. New processes to treat SNF not suitable for disposal in its current form were identified. These processes have several advantages over existing aqueous processes

  1. Receipt capability for foreign research reactor (FRR) spent nuclear fuel (SNF) at the Savannah River Site (SRS)

    Clark, William D. Jr.

    1997-01-01

    The United Stated Department of Energy began implementation of the ten year FRR SNF return policy in May, 1996. Seventeen months into the thirteen year return program, four shipments have been made, returning 863 assemblies of aluminum clad SNF to SRS. Five additional shipments containing over 1,200 assemblies are scheduled in fiscal year 1998. During negotiation of contracts with various reactor operators, it has become apparent that many facilities wish to delay the return of their SNF until the latter part of the program. This has raised concern on the part of the DOE that insufficient receipt capability will exist during the last three to five years of the program to ensure the return of all of the SNF. To help quantify this issue and ensure that it is addressed early in the program, a computer simulation model has been developed at SRS to facilitate the planning, scheduling, and analysis of SNF shipments to be received from offsite facilities. The simulation model, called OFFSHIP, greatly reduces the time and effort required to analyze the complex global transportation system that involves dozens of reactor facilities, multiple casks and fuel types, and time-dependent SNF inventories. OFFSHIP allows the user to input many variables including priorities, cask preferences, shipping date preferences, turnaround times, and regional groupings. User input is easily managed using a spreadsheet format and the output data is generated in a spreadsheet format to facilitate detailed analysis and prepare graphical results. The model was developed in Microsoft Visual Basic for Applications and runs native in Microsoft Excel. The receipt schedules produced by the model have been compared to schedules generated manually with consistent results. For the purposes of this presentation, four scenarios have been developed. The 'Base Case' accounts for those countries/facilities that DOE believes may not participate in the return program. The three additional scenarios look at the

  2. Spent nuclear fuel sampling strategy

    Bergmann, D.W.

    1995-01-01

    This report proposes a strategy for sampling the spent nuclear fuel (SNF) stored in the 105-K Basins (105-K East and 105-K West). This strategy will support decisions concerning the path forward SNF disposition efforts in the following areas: (1) SNF isolation activities such as repackaging/overpacking to a newly constructed staging facility; (2) conditioning processes for fuel stabilization; and (3) interim storage options. This strategy was developed without following the Data Quality Objective (DQO) methodology. It is, however, intended to augment the SNF project DQOS. The SNF sampling is derived by evaluating the current storage condition of the SNF and the factors that effected SNF corrosion/degradation

  3. Acceptance of failed SNF [spent nuclear fuel] assemblies by the Federal Waste Management System

    1990-03-01

    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. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for delivery; and defense and commercial high-level waste packages. This document discusses acceptance of failed spent fuel assemblies by the Federal Waste Management System. 18 refs., 7 figs., 25 tabs

  4. Duo_2-Steel cermet manufacturing technology for PWR Spent Nuclear Fuel (SNF) casks

    Siti Alimah; Budiarto

    2005-01-01

    Assessment of DUO_2-Steel cermet manufacturing technology for PWR SNF casks has been done. DUO_2-Steel cermet consisting of DUO_2 particulates and other particulates, embedded in a steel matrix. Cermet SNF casks have the potential for superior performance compared with casks constructed of other materials. The addition of DUO_2 ceramic particulates can increase SNF cask capacity, improve of repository performance and disposal of excess depleted uranium as potential waste. Two sets of cermet manufacturing technologies are casting and powder metallurgy. Three casting methods are infusion casting, traditional casting and centrifugal casting. While for powder metallurgy methods there are traditional method and new method. DUO_2-Steel cermet have traditionally been produced by powder metallurgy methods. The production of a cask, however, presents special requirements: the manufacture of an annular object with weights up to 100 tons, and methods are being not to manufacture a cermet of this size and geometry. A new powder metallurgy method, is a method for manufacturing cermet for PWR SNF cask. This powder metallurgy techniques have potentials low costs and provides greater freedom In the design of the cermet cask by allowing variable cermet properties. (author)

  5. DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE

    T.L. Mitchell

    2000-01-01

    The purpose of this analysis is to demonstrate the design of the naval spent nuclear fuel (SNF) waste package (WP) using the Waste Package Department's (WPD) design methodologies and processes described in the ''Waste Package Design Methodology Report'' (CRWMS MandO [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000b). The calculations that support the design of the naval SNF WP will be discussed; however, only a sub-set of such analyses will be presented and shall be limited to those identified in the ''Waste Package Design Sensitivity Report'' (CRWMS MandO 2000c). The objective of this analysis is to describe the naval SNF WP design method and to show that the design of the naval SNF WP complies with the ''Naval Spent Nuclear Fuel Disposal Container System Description Document'' (CRWMS MandO 1999a) and Interface Control Document (ICD) criteria for Site Recommendation. Additional criteria for the design of the naval SNF WP have been outlined in Section 6.2 of the ''Waste Package Design Sensitivity Report'' (CRWMS MandO 2000c). The scope of this analysis is restricted to the design of the naval long WP containing one naval long SNF canister. This WP is representative of the WPs that will contain both naval short SNF and naval long SNF canisters. The following items are included in the scope of this analysis: (1) Providing a general description of the applicable design criteria; (2) Describing the design methodology to be used; (3) Presenting the design of the naval SNF waste package; and (4) Showing compliance with all applicable design criteria. The intended use of this analysis is to support Site Recommendation reports and assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the technical product development plan (TPDP) ''Design Analysis for the Naval SNF Waste Package (CRWMS MandO 2000a)

  6. Spent Nuclear Fuel (SNF) Project Cask and MCO Helium Purge System Design Review Completion Report - Project A.5 and A.6

    ARD, K.E.

    2000-01-01

    This report documents the results of the design verification performed on the Cask and Multiple Canister Over-pack (MCO) Helium Purge System. The helium purge system is part of the Spent Nuclear Fuel (SNF) Project Cask Loadout System (CLS) at 100K area. The design verification employed the ''Independent Review Method'' in accordance with Administrative Procedure (AP) EN-6-027-01

  7. Thermal Analysis of Cold Vacuum Drying (CVD) of Spent Nuclear Fuel (SNF)

    PIEPHO, M.G.

    2000-01-01

    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 N Reactor spent fuel. This analysis provides the basis for the MCO thermal behavior at the CVD Facility in support of the safety basis documentation

  8. Characterization of FRR SNF in Basin and Dry Storage Systems

    Brooks, H.M.; Sindelar, R.L.

    1998-09-01

    Since May 1996, over 1700 aluminum-based spent nuclear fuel (A1-SNF) assemblies have been inspected for corrosion and mechanical damage to determine if the cladding had been penetrated as part of the process for acceptance of the fuel at the Savannah River Site (SRS). The results of the release measurements are summarized in this paper

  9. Commercial SNF Accident Release Fractions

    J. Schulz

    2004-11-05

    The purpose of this analysis is to specify and document the total and respirable fractions for radioactive materials that could be potentially released from an accident at the repository involving commercial spent nuclear fuel (SNF) in a dry environment. The total and respirable release fractions are used to support the preclosure licensing basis for the repository. The total release fraction is defined as the fraction of total commercial SNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. Radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses; this subset of the total release fraction is referred to as the respirable release fraction. Accidents may involve waste forms characterized as: (1) bare unconfined intact fuel assemblies, (2) confined intact fuel assemblies, or (3) canistered failed commercial SNF. Confined intact commercial SNF assemblies at the repository are contained in shipping casks, canisters, or waste packages. Four categories of failed commercial SNF are identified: (1) mechanically and cladding-penetration damaged commercial SNF, (2) consolidated/reconstituted assemblies, (3) fuel rods, pieces, and debris, and (4) nonfuel components. It is assumed that failed commercial SNF is placed into waste packages with a mesh screen at each end (CRWMS M&O 1999). In contrast to bare unconfined fuel assemblies, the

  10. Commercial SNF Accident Release Fractions

    Schulz, J.

    2004-01-01

    The purpose of this analysis is to specify and document the total and respirable fractions for radioactive materials that could be potentially released from an accident at the repository involving commercial spent nuclear fuel (SNF) in a dry environment. The total and respirable release fractions are used to support the preclosure licensing basis for the repository. The total release fraction is defined as the fraction of total commercial SNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. Radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses; this subset of the total release fraction is referred to as the respirable release fraction. Accidents may involve waste forms characterized as: (1) bare unconfined intact fuel assemblies, (2) confined intact fuel assemblies, or (3) canistered failed commercial SNF. Confined intact commercial SNF assemblies at the repository are contained in shipping casks, canisters, or waste packages. Four categories of failed commercial SNF are identified: (1) mechanically and cladding-penetration damaged commercial SNF, (2) consolidated/reconstituted assemblies, (3) fuel rods, pieces, and debris, and (4) nonfuel components. It is assumed that failed commercial SNF is placed into waste packages with a mesh screen at each end (CRWMS M andO 1999). In contrast to bare unconfined fuel assemblies, the

  11. A method for selection of spent nuclear fuel (SNF) transportation route considering socioeconomic cost based on contingent valuation method (CVM)

    Kim, Young Sik

    2008-02-01

    A transportation of SNF may cause an additional radiation exposure to human beings. It means that the radiological risk should be estimated and managed quantitatively for the public who live near the shipments route. Before the SNF transportation is performed, the route selection is concluded based on the radiological risk estimated with RADTRAN code in existing method generally. It means the existing method for route selection is based only on the radiological health risk but there are not only the impacts related to the radiological health risk but also the socioeconomic impacts related to the cost. In this study, a new method and its numerical formula for route selection on transporting SNF is proposed based on cost estimation because there are several costs in transporting SNF. The total cost consists of radiological health cost, transportation cost, and socioeconomic cost. Each cost is defined properly to the characteristics of SNF transportation and many coefficients and variables describing the meaning of each cost are obtained or estimated through many surveys. Especially to get the socioeconomic cost, contingent valuation method (CVM) is used with a questionnaire. The socioeconomic cost estimation is the most important part of the total cost originated from transporting SNF because it is a very dominant cost in the total cost. The route selection regarding SNF transportation can be supported with the proposed method reasonably and unnecessary or exhausting controversies about the shipments could be avoided

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

    MITCHELL, R.M.

    2000-01-01

    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

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

    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.

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

    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.

  15. Idaho Chemical Processing Plant Spent Fuel and Waste Management Technology Development Program Plan

    1993-09-01

    The Department of Energy (DOE) has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage and reprocessing since 1953. Reprocessing of SNF has resulted in an existing inventory of 1.5 million gallons of radioactive sodium-bearing liquid waste and 3800 cubic meters (m 3 ) of calcine, in addition to the 768 metric tons (MT) of SNF and various other fuel materials in inventory. To date, the major activity of the ICPP has been the reprocessing of SNF to recover fissile uranium; however, recent changes in world events have diminished the demand to recover and recycle this material. As a result, DOE has discontinued reprocessing SNF for uranium recovery, making the need to properly manage and dispose of these and future materials a high priority. In accordance with the Nuclear Waste Policy Act (NWPA) of 1982, as amended, disposal of SNF and high-level waste (HLW) is planned for a geological repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP Spent Fuel and Waste Management Technology Development Program (SF ampersand WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will properly stored and prepared for final disposal. Program elements in support of acceptable interim storage and waste minimization include: developing and implementing improved radioactive waste treatment technologies; identifying and implementing enhanced decontamination and decommissioning techniques; developing radioactive scrap metal (RSM) recycle capabilities; and developing and implementing improved technologies for the interim storage of SNF

  16. Idaho Chemical Processing Plant spent fuel and waste management technology development program plan: 1994 Update

    1994-09-01

    The Department of Energy has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage since 1951 and reprocessing since 1953. Until April 1992, the major activity of the ICPP was the reprocessing of SNF to recover fissile uranium and the management of the resulting high-level wastes (HLW). In 1992, DOE chose to discontinue reprocessing SNF for uranium recovery and shifted its focus toward the continued safe management and disposition of SNF and radioactive wastes accumulated through reprocessing activities. Currently, 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3,800 cubic meters of calcine waste, and 289 metric tons heavy metal of SNF are in inventory at the ICPP. Disposal of SNF and high-level waste (HLW) is planned for a repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP spent Fuel and Waste Management Technology Development Program (SF ampersand WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will be properly stored and prepared for final disposal in accordance with regulatory drivers. This Plan presents a brief summary of each of the major elements of the SF ampersand WMTDP; identifies key program assumptions and their bases; and outlines the key activities and decisions that must be completed to identify, develop, demonstrate, and implement a process(es) that will properly prepare the SNF and radioactive wastes stored at the ICPP for safe and efficient interim storage and final disposal

  17. Fuel processing. Wastes processing

    Bourgeois, M.

    2000-01-01

    The gaseous, liquid and solid radioactive effluents generated by the fuel reprocessing, can't be release in the environment. They have to be treated in order to respect the limits of the pollution regulations. These processing are detailed and discussed in this technical paper. A second part is devoted to the SPIN research program relative to the separation of the long life radionuclides in order to reduce the radioactive wastes storage volume. (A.L.B.)

  18. Spent nuclear fuel project product specification

    PAJUNEN, A.L.

    1999-01-01

    This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project

  19. Spent Nuclear Fuel project stage and store K basin SNF in canister storage building functions and requirements. Revision 1

    Womack, J.C.

    1995-01-01

    This document establishes the functions and requirements baseline for the implementation of the Canister Storage Building Subproject. The mission allocated to the Canister Storage Building Subproject is to provide safe, environmentally sound staging and storage of K Basin SNF until a decision on the final disposition is reached and implemented

  20. A direct, single-step plasma arc-vitreous ceramic process for stabilizing spent nuclear fuels, sludges, and associated wastes

    Feng, X.; Einziger, R.E.; Eschenbach, R.C.

    1997-01-01

    A single-step plasma arc-vitreous ceramic (PAVC) process is described for converting spent nuclear fuel (SNF), SNF sludges, and associated wastes into a vitreous ceramic waste form. This proposed technology is built on extensive experience of nuclear waste form development and nuclear waste treatment using the commercially available plasma arc centrifugal (PAC) system. SNF elements will be loaded directly into a PAC furnace with minimum additives and converted into vitreous ceramics with up to 90 wt% waste loading. The vitreous ceramic waste form should meet the functional requirements for borosilicate glasses for permanent disposal in a geologic repository and for interim storage. Criticality safety would be ensured through the use of batch modes, and controlling the amount of fuel processed in one batch. The minimum requirements on SNF characterization and pretreatment, the one-step process, and minimum secondary waste generation may reduce treatment duration, radiation exposure, and treatment cost

  1. Nuclear fuels

    Gangwani, Saloni; Chakrabortty, Sumita

    2011-01-01

    Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

  2. Canister storage building compliance assessment SNF project NRC equivalency criteria - HNF-SD-SNF-DB-003

    BLACK, D.M.

    1999-01-01

    This document presents the Project's position on compliance with the SNF Project NRC Equivalency Criteria - HNF-SD-SNF-DE-003, Spent Nuclear Fuel Project Path Forward Additional NRC Requirements. No non-compliances are shown. The compliance statements have been reviewed and approved by DOE. Open items are scheduled to be closed prior to project completion

  3. Probability of Criticality for MOX SNF

    P. Gottlieb

    1999-01-01

    The purpose of this calculation is to provide a conservative (upper bound) estimate of the probability of criticality for mixed oxide (MOX) spent nuclear fuel (SNF) of the Westinghouse pressurized water reactor (PWR) design that has been proposed for use. with the Plutonium Disposition Program (Ref. 1, p. 2). This calculation uses a Monte Carlo technique similar to that used for ordinary commercial SNF (Ref. 2, Sections 2 and 5.2). Several scenarios, covering a range of parameters, are evaluated for criticality. Parameters specifying the loss of fission products and iron oxide from the waste package are particularly important. This calculation is associated with disposal of MOX SNF

  4. Actual Situation and Further Development of Interim Storage of Spent Nuclear Fuel (SNF) and Highly Active Waste (HAW) from the View of the Competent Authority in the Field of section 6

    Gastl, Christoph; Drobniewski, Christian

    2014-01-01

    According to the German atomic law the storage of nuclear material has to be licensed following section 6 by the competent authority in this field, which is the Federal Office for Radiation Protection. Interim storage in its actual form started in 2002 in the interim storage facility next to the NPP Lingen. Since this time each NPP erected its own storage facilities and three central storage facilities have been built. The spent nuclear fuel (SNF) and the vitrified high level waste (HAW) will be stored there until final disposal. The time span from now on to the point of opening of a final disposal facility shall be presented from a regulators point of view, divided into different phase which could spread from years to decades. Special attention shall be drawn on the different aspects influencing the licensing process and its duration at the moment and in future including the capabilities of the competent authority. (authors)

  5. Technical Approach and Plan for Transitioning Spent Nuclear Fuel (SNF) Project Facilities to the Environmental Restoration Program

    SKELLY, W.A.

    1999-01-01

    This document describes the approach and process in which the 100-K Area Facilities are to be deactivated and transitioned over to the Environmental Restoration Program after spent nuclear fuel has been removed from the K Basins. It describes the Transition Project's scope and objectives, work breakdown structure, activity planning, estimated cost, and schedule. This report will be utilized as a planning document for project management and control and to communicate details of project content and integration

  6. Fuels Processing Laboratory

    Federal Laboratory Consortium — NETL’s Fuels Processing Laboratory in Morgantown, WV, provides researchers with the equipment they need to thoroughly explore the catalytic issues associated with...

  7. Fuels processing for transportation fuel cell systems

    Kumar, R.; Ahmed, S.

    Fuel cells primarily use hydrogen as the fuel. This hydrogen must be produced from other fuels such as natural gas or methanol. The fuel processor requirements are affected by the fuel to be converted, the type of fuel cell to be supplied, and the fuel cell application. The conventional fuel processing technology has been reexamined to determine how it must be adapted for use in demanding applications such as transportation. The two major fuel conversion processes are steam reforming and partial oxidation reforming. The former is established practice for stationary applications; the latter offers certain advantages for mobile systems and is presently in various stages of development. This paper discusses these fuel processing technologies and the more recent developments for fuel cell systems used in transportation. The need for new materials in fuels processing, particularly in the area of reforming catalysis and hydrogen purification, is discussed.

  8. SNF shipping cask shielding analysis

    Johnson, J.O.; Pace, J.V. III.

    1996-01-01

    The Waste Management and Remedial Action Division has planned a modification sequence for storage facility 7827 in the Solid Waste Storage Area (SWSA). The modification cycle is: (1) modify an empty caisson, (2) transfer the spent nuclear fuel (SNF) of an occupied caisson to a hot cell in building 3525 for inspection and possible repackaging, and (3) return the package to the modified caisson in the SWSA. Although the SNF to be moved is in the solid form, it has different levels of activity. Thus, the following 5 shipping casks will be available for the task: the Loop Transport Carrier, the In- Pile Loop LITR HB-2 Carrier, the 6.5-inch HRLEL Carrier, the HFIR Hot Scrap Carrier, and the 10-inch ORR Experiment Removal Shield Cask. This report describes the shielding tasks for the 5 casks: determination of shielding characteristics, any streaming avenues, estimation of thermal limits, and shielding calculational uncertainty for use in the transportation plan

  9. SNF AGING SYSTEM DESCRIPTION DOCUMENT

    L.L. Swanson

    2005-01-01

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF) aging system and associated bases, which will allow the design effort to proceed. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in the SDD reflects the current results of the design process. Throughout this SDD, the term aging cask applies to vertical site-specific casks and to horizontal aging modules. The term overpack is a vertical site-specific cask that contains a dual-purpose canister (DPC) or a disposable canister. Functional and operational requirements applicable to this system were obtained from ''Project Functional and Operational Requirements'' (F andOR) (Curry 2004 [DIRS 170557]). Other requirements that support the design process were taken from documents such as ''Project Design Criteria Document'' (PDC) (BSC 2004 [DES 171599]), ''Site Fire Hazards Analyses'' (BSC 2005 [DIRS 172174]), and ''Nuclear Safety Design Bases for License Application'' (BSC 2005 [DIRS 171512]). The documents address requirements in the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]). This SDD includes several appendices. Appendix A is a Glossary; Appendix B is a list of key system charts, diagrams, drawings, lists and additional supporting information; and Appendix C is a list of

  10. Implementation process and deployment initiatives for the regionalized storage of DOE-owned spent nuclear fuel

    Dearien, J.A.; Smith, N.E.L.

    1995-01-01

    This report describes how DOE-owned spent nuclear fuel (SNF) will be stored in the interim 40-year period from 1996 to 2035, by which time it is expected to be in a National Nuclear Repository. The process is described in terms of its primary components: fuel inventory, facilities where it is stored, how the fuel will be moved, and legal issues associated with the process. Tools developed to deploy and fulfill the implementation needs of the National Spent Nuclear Fuel Program are also discussed

  11. Comparison of spent nuclear fuel management alternatives

    Beebe, C.L.; Caldwell, M.A.

    1996-01-01

    This paper reports the process an results of a trade study of spent nuclear fuel (SNF)management alternatives. The purpose of the trade study was to provide: (1) a summary of various SNF management alternatives, (2) an objective comparison of the various alternatives to facilitate the decision making process, and (3) documentation of trade study rational and the basis for decisions

  12. Measurements of Fundamental Fluid Physics of SNF Storage Canisters

    Condie, Keith Glenn; Mc Creery, Glenn Ernest; McEligot, Donald Marinus

    2001-09-01

    With the University of Idaho, Ohio State University and Clarksean Associates, this research program has the long-term goal to develop reliable predictive techniques for the energy, mass and momentum transfer plus chemical reactions in drying / passivation (surface oxidation) operations in the transfer and storage of spent nuclear fuel (SNF) from wet to dry storage. Such techniques are needed to assist in design of future transfer and storage systems, prediction of the performance of existing and proposed systems and safety (re)evaluation of systems as necessary at later dates. Many fuel element geometries and configurations are accommodated in the storage of spent nuclear fuel. Consequently, there is no one generic fuel element / assembly, storage basket or canister and, therefore, no single generic fuel storage configuration. One can, however, identify generic flow phenomena or processes which may be present during drying or passivation in SNF canisters. The objective of the INEEL tasks was to obtain fundamental measurements of these flow processes in appropriate parameter ranges.

  13. Fusion reactor fuel processing

    Johnson, E.F.

    1972-06-01

    For thermonuclear power reactors based on the continuous fusion of deuterium and tritium the principal fuel processing problems occur in maintaining desired compositions in the primary fuel cycled through the reactor, in the recovery of tritium bred in the blanket surrounding the reactor, and in the prevention of tritium loss to the environment. Since all fuel recycled through the reactor must be cooled to cryogenic conditions for reinjection into the reactor, cryogenic fractional distillation is a likely process for controlling the primary fuel stream composition. Another practical possibility is the permeation of the hydrogen isotopes through thin metal membranes. The removal of tritium from the ash discharged from the power system would be accomplished by chemical procedures to assure physiologically safe concentration levels. The recovery process for tritium from the breeder blanket depends on the nature of the blanket fluids. For molten lithium the only practicable possibility appears to be permeation from the liquid phase. For molten salts the process would involve stripping with inert gas followed by chemical recovery. In either case extremely low concentrations of tritium in the melts would be desirable to maintain low tritium inventories, and to minimize escape of tritium through unwanted permeation, and to avoid embrittlement of metal walls. 21 refs

  14. Peculiarities of highly burned-up NPP SNF reprocessing and new approach to simulation of solvent extraction processes

    Fedorov, Y.S.; Zilberman, B.Y.; Goletskiy, N.D.; Puzikov, E.A.; Ryabkov, D.V.; Rodionov, S.A.; Beznosyuk, V.I.; Petrov, Y.Y.; Saprykin, V.F.; Murzin, A.A.; Bibichev, B.A.; Aloy, A.S.; Kudinov, A.S.; Blazheva, I.V. [RPA ' V.G.Khlopin Radium Institute' , 28, 2 Murinsky av., St-Petersburg, 194 021 (Russian Federation); Kurenkov, N.V. [Institute of Industrial Nuclear Technology NRNU MEPHI, 31, Kashirskoye shosse, Moscow, 115409 (Russian Federation)

    2013-07-01

    Substantiation, general description and performance characteristics of a reprocessing flowsheet for WWER-1000 spent fuel with burn-up >60 GW*day/t U is given. Pu and U losses were <0.1%, separation factor > 10{sup 4}; their decontamination factor from γ-emitting fission products was 4*10{sup 4} and 3*10{sup 7}, respectively. Zr, Tc, Np removal was >98% at U and Pu losses <0.05%. A new approach to simulation of extraction equilibrium has been developed. It is based on a set of simultaneous chemical reactions characterized by apparent concentration constants. A software package was created for simulation of spent fuel component distribution in multistage countercurrent extraction processes in the presence of salting out agents. (authors)

  15. SNF AGING SYSTEM DESCRIPTION DOCUMENT

    L.L. Swanson

    2005-04-06

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF) aging system and associated bases, which will allow the design effort to proceed. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flow down of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in the SDD reflects the current results of the design process. Throughout this SDD, the term aging cask applies to vertical site-specific casks and to horizontal aging modules. The term overpack is a vertical site-specific cask that contains a dual-purpose canister (DPC) or a disposable canister. Functional and operational requirements applicable to this system were obtained from ''Project Functional and Operational Requirements'' (F&OR) (Curry 2004 [DIRS 170557]). Other requirements that support the design process were taken from documents such as ''Project Design Criteria Document'' (PDC) (BSC 2004 [DES 171599]), ''Site Fire Hazards Analyses'' (BSC 2005 [DIRS 172174]), and ''Nuclear Safety Design Bases for License Application'' (BSC 2005 [DIRS 171512]). The documents address requirements in the ''Project Requirements Document'' (PRD) (Canori and Leitner 2003 [DIRS 166275]). This SDD includes several appendices. Appendix A is a Glossary; Appendix B is a list of key system charts

  16. Current status of development in dry pyro-electrochemical technology of SNF reprocessing

    Bychkov, A.V.; Skiba, O.V.; Kormilitsyn, M.V.

    2004-01-01

    The technology of SNF management in molten salts currently developed by a group of institutes headed by RIAR has had several stages of development: - basic research of uranium, plutonium and main FP properties (investigation and reprocessing of different kinds of SNF in 1960 - 1970); - development of the equipment and implementation of the pyro-electrochemical technology of granulated UPu fuel production. Development of the vibro-packing method and in-pile testing of vibro-packed fuel pins with granulated fuel as the most 'logical' continuation of reprocessing: implementation of the technology for BOR-60 and BN-600 (1980 - 1990); - development of closed fuel cycle elements. Checking of the technology using batches of SNF. In-pile tests. Feasibility study of the closed fuel cycle (CFC). Study of application of the technology to other objects (transmutation; nitride, cermet and other fuels) (1980 - 1990). The current status of the research is the following: - Basic research. Properties of uranium, plutonium, thorium, and neptunium in chloride melts have been studied in much detail. The data on physical chemistry and electrochemistry of the main FP is enough for understanding the processes. Detailed studies of americium, curium, and technetium chemistry are the essential investigation directions; - Engineering development. The technology and equipment bases have been developed for the processes of oxide fuel reprocessing and fabrication. The technology was checked using 5500 kg of pure fuel from different reactors and 20 kg of irradiated BN-350 and BOR-60 fuel. The bases of the technology have been provided and the feasibility study has been carried out for a full-scale plant of BN-800 CFC; - Industrial application: Since the technology is highly prepared, the activities on industrial application of U-Pu fuel are now underway. The BOR-60 reactor uses fuel obtained by the dry method, the design of the facility for implementation of CFC reactors is being developed. 9

  17. SPENT NUCLEAR FUEL (SNF) PROJECT CANISTER STORAGE BUILDING (CSB) MULTI CANISTER OVERPACK (MCO) SAMPLING SYSTEM VALIDATION (OCRWM)

    BLACK, D.M.; KLEM, M.J.

    2003-01-01

    Approximately 400 Multi-canister overpacks (MCO) containing spent nuclear fuel are to be interim stored at the Canister Storage Building (CSB). Several MCOs (monitored MCOs) are designated to be gas sampled periodically at the CSB sampling/weld station (Bader 2002a). The monitoring program includes pressure, temperature and gas composition measurements of monitored MCOs during their first two years of interim storage at the CSB. The MCO sample cart (CART-001) is used at the sampling/weld station to measure the monitored MCO gas temperature and pressure, obtain gas samples for laboratory analysis and refill the monitored MCO with high purity helium as needed. The sample cart and support equipment were functionally and operationally tested and validated before sampling of the first monitored MCO (H-036). This report documents the results of validation testing using training MCO (TR-003) at the CSB. Another report (Bader 2002b) documents the sample results from gas sampling of the first monitored MCO (H-036). Validation testing of the MCO gas sampling system showed the equipment and procedure as originally constituted will satisfactorily sample the first monitored MCO. Subsequent system and procedural improvements will provide increased flexibility and reliability for future MCO gas sampling. The physical operation of the sampling equipment during testing provided evidence that theoretical correlation factors for extrapolating MCO gas composition from sample results are unnecessarily conservative. Empirically derived correlation factors showed adequate conservatism and support use of the sample system for ongoing monitored MCO sampling

  18. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D.

    2013-01-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  19. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  20. SNF/HLW Transfer System Description Document

    W. Holt

    2005-01-01

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF)/high-level radioactive waste (HLW) transfer system and associated bases, which will allow the design effort to proceed to license application. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control. Accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flowdown of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in this SDD reflects the current results of the design process

  1. SNF sludge treatment system preliminary project execution plan

    Flament, T.A.

    1998-01-01

    The Fluor Daniel Hanford, Inc. (FDH) Project Director for the Spent Nuclear Fuel (SNF) Project has requested Numatec Hanford Company (NHC) to define how Hanford would manage a new subproject to provide a process system to receive and chemically treat radioactive sludge currently stored in the 100 K Area fuel retention basins. The subproject, named the Sludge Treatment System (STS) Subproject, provides and operates facilities and equipment to chemically process K Basin sludge to meet Tank Waste Remediation System (TWRS) requirements. This document sets forth the NHC management approach for the STS Subproject and will comply with the requirements of the SNF Project Management Plan (HNF-SD-SNFPMP-011). This version of this document is intended to apply to the initial phase of the subproject and to evolve through subsequent revision to include all design, fabrication, and construction conducted on the project and the necessary management and engineering functions within the scope of the subproject. As Project Manager, NHC will perform those activities necessary to complete the STS Subproject within approved cost and schedule baselines and turn over to FDH facilities, systems, and documentation necessary for operation of the STS

  2. Compound process fuel cycle concept

    Ikegami, Tetsuo

    2005-01-01

    Mass flow of light water reactor spent fuel for a newly proposed nuclear fuel cycle concept 'Compound Process Fuel Cycle' has been studied in order to assess the capacity of the concept for accepting light water reactor spent fuels, taking an example for boiling water reactor mixed oxide spent fuel of 60 GWd/t burn-up and for a fast reactor core of 3 GW thermal output. The acceptable heavy metal of boiling water reactor mixed oxide spent fuel is about 3.7 t/y/reactor while the burn-up of the recycled fuel is about 160 GWd/t and about 1.6 t/y reactor with the recycled fuel burn-up of about 300 GWd/t, in the case of 2 times recycle and 4 times recycle respectively. The compound process fuel cycle concept has such flexibility that it can accept so much light water reactor spent fuels as to suppress the light water reactor spent fuel pile-up if not so high fuel burn-up is expected, and can aim at high fuel burn-up if the light water reactor spent fuel pile-up is not so much. Following distinctive features of the concept have also been revealed. A sort of ideal utilization of boiling water reactor mixed oxide spent fuel might be achieved through this concept, since both plutonium and minor actinide reach equilibrium state beyond 2 times recycle. Changes of the reactivity coefficients during recycles are mild, giving roughly same level of reactivity coefficients as the conventional large scale fast breeder core. Both the radio-activity and the heat generation after 4 year cooling and after 4 times recycle are less than 2.5 times of those of the pre recycle fuel. (author)

  3. International topical meeting on research reactor fuel management (RRFM) - United States foreign research reactor (FRR) spent nuclear fuel (SNF) acceptance program: 2010 update

    Messick, C.E.; Taylor, J.L.; Niehus, M.T.; Landers, C.

    2010-01-01

    The Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, adopted by the United States Department of Energy (DOE), in consultation with the Department of State (DOS) in May 1996, scheduled to expire May 12, 2016, to return research reactor fuel until May 12, 2019 to the U.S. is in its fourteenth year. This paper provides a brief update on the program, part of the National Nuclear Security Administration (NNSA), and discusses program initiatives and future activities. The goal of the program continues to be recovery of U.S.-origin nuclear materials, which could otherwise be used in weapons, while assisting other countries to enjoy the benefits of nuclear technology. The NNSA is seeking feedback from research reactor operators to help us understand ways to include eligible research reactors who have not yet participated in the program. (author)

  4. International topical meeting on research reactor fuel management (RRFM) - United States Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) acceptance program: 2007 update

    Messick, C.E.; Taylor, J.L.

    2007-01-01

    The Nuclear Weapons Non-proliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, adopted by The United States Department of Energy (DOE), in consultation with the Department of State in May 1996, has been extended to expire May 12, 2016, providing an additional 10 years to return fuel to the U.S. This paper provides a brief update on the program, now transferred to the National Nuclear Security Administration (NNSA), and discusses program initiatives and future activities. The goal of the program continues to be recovery of nuclear materials (27 countries have participated so far, returning a total of 7620 spent nuclear fuel elements), which could otherwise be used in weapons, while assisting other countries to enjoy the benefits of nuclear technology. More than ever before, DOE and reactor operators need to work together to schedule shipments as soon as possible, to optimize shipment efficiency over the remaining years of the program. The NNSA is seeking feedback from research reactor operators to help us understand ways to include eligible reactor who have not yet participated in the program

  5. Vacuum Drying Tests for Storage of Aluminum Spent Nuclear Fuel

    Chen, K.F.; Large, W.S.; Sindelar, R.L.

    1998-05-01

    A total inventory of up to approximately 32,000 aluminum-based spent nuclear fuel (Al SNF) assemblies are expected to be shipped to Savannah River Site (SRS) from domestic and foreign research reactors over the next several decades. Treatment technologies are being developed as alternatives to processing for the ultimate disposition of Al SNF in the geologic repository. One technology, called Direct/Co-disposal of Al SNF, would place the SNF into a canister ready for disposal in a waste package, with or without canisters containing high-level radioactive waste glass logs, in the repository. The Al SNF would be transferred from wet storage and would need to be dried in the Al SNF canister. The moisture content inside the Al SNF canister is limited to avoid excessive Al SNF corrosion and hydrogen buildup during interim storage before disposal. A vacuum drying process was proposed to dry the Al SNF in a canister. There are two major concerns for the vacuum drying process. One is water inside the canister could become frozen during the vacuum drying process and the other one is the detection of dryness inside the canister. To vacuum dry an irradiated fuel in a heavily shielded canister, it would be very difficult to open the lid to inspect the dryness during the vacuum drying operation. A vacuum drying test program using a mock SNF assembly was conducted to demonstrate feasibility of drying the Al SNF in a canister. These tests also served as a check-out of the drying apparatus for future tests in which irradiated fuel would be loaded into a canister under water followed by drying for storage

  6. CONTAINMENT EVALUATION OF BREACHED AL-SNF FOR CASK TRANSPORT

    Vinson, D. W.; Sindelar, R. L.; Iyer, N. C.

    2005-01-01

    Aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site. To enter the U.S., the cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Al-SNF is subject to corrosion degradation in water storage, and many of the fuel assemblies are ''failed'' or have through-clad damage. A methodology has been developed with technical bases to show that Al-SNF with cladding breaches can be directly transported in standard casks and maintained within the allowable release rates. The approach to evaluate the limiting allowable leakage rate, L R , for a cask with breached Al-SNF for comparison to its test leakage rate could be extended to other nuclear material systems. The approach for containment analysis of Al-SNF follows calculations for commercial spent fuel as provided in NUREG/CR-6487 that adopts ANSI N14.5 as a methodology for containment analysis. The material-specific features and characteristics of damaged Al-SNF (fuel materials, fabrication techniques, microstructure, radionuclide inventory, and vapor corrosion rates) that were derived from literature sources and/or developed in laboratory testing are applied to generate the four containment source terms that yield four separate cask cavity activity densities; namely, those from fines; gaseous fission product species; volatile fission product species; and fuel assembly crud. The activity values, A 2 , are developed per the guidance of 10CFR71. The analysis is performed parametrically to evaluate maximum number of breached assemblies and exposed fuel area for a proposed shipment in a cask with a test leakage rate

  7. Evaluation of Test Methodologies for Dissolution and Corrosion of Al-SNF

    Wiersma, B.J.; Mickalonis, J.I.; Louthan, M.R.

    1998-09-01

    The performance of aluminum-based spent nuclear fuel (Al-SNF) in the repository will differ from that of the commercial nuclear fuels and the high level waste glasses. The program consists of evaluating three test methods

  8. Catalytic processes for cleaner fuels

    Catani, R.; Marchionna, M.; Rossini, S.

    1999-01-01

    More stringent limitations on vehicle emissions require different measurement: fuel reformulation is one of the most important and is calling for a noticeable impact on refinery assets. Composition rangers of the future fuels have been defined on a time scale. In this scenario the evolution of catalytic technologies becomes a fundamental tool for allowing refinery to reach the fixed-by-law targets. In this paper, the refinery process options to meet each specific requirements of reformulated fuels are surveyed [it

  9. Legal basis of Russian origin irradiated WWER nuclear fuel import to the Russian Federation

    Kanashov, B.; Dorofeev, A.; Komarov, S.; Smirnov, V.; Kolupaev, D.; Kriger, O.

    2008-01-01

    In the process of spent nuclear fuel (SNF) returning from Armenia, Bulgaria, Hungary, Slovakia, Ukraine, Finland and the Czech Republic to the Russian Federation the following issues have to be considered: 1) Does the legal opportunity of SNF import to the RF exist? 2) Does the technical opportunity for SNF acceptance at reprocessing or disposal facility exist? 3) What are the basic conditions for SNF import? 4) What are the basic conditions for return or retaining of reprocessing products including RAW? The first issue is a legal one and has to be resolved within the framework of federal laws, RF government regulations and international agreements. The second issue is normative-technical. It is regulated by documents of Rostechnadzor (Federal agency on ecological, technological and nuclear supervision), federal norms and regulations in the field of atomic energy usage, industry standards, and in case they are absent, by technical specifications for SNF supply. The last two issues are resolved in the process of drafting foreign trade contracts on SNF import. Generally, Russian regulatory framework is developed enough to regulate SNF import and handling, even in most complicated cases. Nevertheless, when foreign trade contracts on SNF import being drafted there may be disputed regarding both SNF import and RAW return. This report concerns the RF legal and regulatory basis on terms and conditions of SNF import, interim storage, reprocessing and reprocessing products handling in the RF. (authors)

  10. Development of the ENVI simulator to estimate Korean SNF flow and its cost - 16060

    Hwang, Yongsoo; Miller, Ian

    2009-01-01

    This paper describes an integrated model developed by the Korean Atomic Energy Research Institute (KAERI) to simulate options for managing spent nuclear fuel (SNF) in South Korea. A companion paper (Hwang and Miller, 2009) describes a performance assessment model to address the long-term safety of alternative geological disposal options for different waste streams. The model addresses alternative concepts for storage, transportation, and processing of SNF of different types (Candu, PWR), leading up to permanent disposal in geological repositories. It uses the GoldSim software to simulate the logistics of the associated activities, including the associated capital and operating costs. The model's results allow direct comparison of alternative waste management concepts, and predict the sizes and timings of different facilities required. Future versions of the model will also address the uncertainties associated with the different system components in order to provide risk-based assessments. (authors)

  11. An assessment of KW Basin radionuclide activity when opening SNF canisters

    Bergmann, D.W.; Mollerus, F.J.; Wray, J.L.

    1995-01-01

    N Reactor spent fuel is being stored in sealed canisters in the KW Basin. Some of the canisters contain damaged fuel elements. There is the potential for release of Cs 137, Kr 85, H3, and other fission products and transuranics (TRUs) when canisters are opened. Canister opening is required to select and transfer fuel elements to the 300 Area for examination as part of the Spent Nuclear Fuel (SNF) Characterization program. This report estimates the amount of radionuclides that can be released from Mark II spent nuclear fuel (SNF) canisters in KW Basin when canisters are opened for SNF fuel sampling as part of the SNF Characterization Program. The report also assesses the dose consequences of the releases and steps that can be taken to reduce the impacts of these releases

  12. Potential dispositioning flowsheets for ICPP SNF and wastes

    Olson, A.L. [ed.; Anderson, P.A.; Bendixsen, C.L. [and others

    1995-11-01

    The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1953. This activity resulted mainly in the recovery of uranium and the management of the resulting wastes. The acidic radioactive high-level liquid waste was routinely stored in stainless steel tanks and then calcined to form a dry granular solid. The calcine is stored in stainless steel bins that are housed in underground concrete vaults. In April 1992, the DOE discontinued the practice of reprocessing irradiated nuclear fuels. This decision has left a legacy of 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3800 cubic meters of calcine waste, and 289 metric tons of heavy metal within unprocessed spent nuclear fuel (SNF) left in inventory at the ICPP. The nation`s radioactive waste policy has been established by the Nuclear Waste Policy Act (NWPA), which requires the final disposal of SNF and radioactive waste in accordance with US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) standards. In accordance with these regulations and other legal agreements between the State of Idaho and the DOE, the DOE must, among other requirements, (1) complete a final Environmental Impact Statement by April 30, 1995, (2) evaluate and test sodium-bearing waste pre-treatment technologies, (3) select the sodium-bearing and calcine waste pre-treatment technology, if necessary, by June 1, 1995, and (4) select a technology for converting calcined waste into an appropriate disposal form by June 1, 1995.

  13. Potential dispositioning flowsheets for ICPP SNF and wastes

    Olson, A.L.; Anderson, P.A.; Bendixsen, C.L.

    1995-11-01

    The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1953. This activity resulted mainly in the recovery of uranium and the management of the resulting wastes. The acidic radioactive high-level liquid waste was routinely stored in stainless steel tanks and then calcined to form a dry granular solid. The calcine is stored in stainless steel bins that are housed in underground concrete vaults. In April 1992, the DOE discontinued the practice of reprocessing irradiated nuclear fuels. This decision has left a legacy of 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3800 cubic meters of calcine waste, and 289 metric tons of heavy metal within unprocessed spent nuclear fuel (SNF) left in inventory at the ICPP. The nation's radioactive waste policy has been established by the Nuclear Waste Policy Act (NWPA), which requires the final disposal of SNF and radioactive waste in accordance with US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) standards. In accordance with these regulations and other legal agreements between the State of Idaho and the DOE, the DOE must, among other requirements, (1) complete a final Environmental Impact Statement by April 30, 1995, (2) evaluate and test sodium-bearing waste pre-treatment technologies, (3) select the sodium-bearing and calcine waste pre-treatment technology, if necessary, by June 1, 1995, and (4) select a technology for converting calcined waste into an appropriate disposal form by June 1, 1995

  14. Evaluation of Neutron Poison Materials for DOE SNF Disposal Systems

    Vinson, D.W.; Caskey, G.R. Jr.; Sindelar, R.L.

    1998-09-01

    Aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors is being consolidated at the Savannah River Site (SRS) for ultimate disposal in the Mined Geologic Disposal System (MGDS). Most of the aluminum-based fuel material contains highly enriched uranium (HEU) (more than 20 percent 235U), which challenges the preclusion of criticality events for disposal periods exceeding 10,000 years. Recent criticality analyses have shown that the addition of neutron absorbing materials (poisons) is needed in waste packages containing DOE SNF canisters fully loaded with Al-SNF under flooded and degraded configurations to demonstrate compliance with the requirement that Keff less than 0.95. Compatibility of poison matrix materials and the Al-SNF, including their relative degradation rate and solubility, are important to maintain criticality control. An assessment of the viability of poison and matrix materials has been conducted, and an experimental corrosion program has been initiated to provide data on degradation rates of poison and matrix materials and Al-SNF materials under repository relevant vapor and aqueous environments. Initial testing includes Al6061, Type 316L stainless steel, and A516Gr55 in synthesized J-13 water vapor at 50 degrees C, 100 degrees C, and 200 degrees C and in condensate water vapor at 100 degrees C. Preliminary results are presented herein

  15. Interaction of DOE SNF and Packaging Materials

    Anderson, P.A.

    1998-01-01

    A sensitivity analysis was conducted to identify and evaluate potential destructive interactions between the materials in US Department of Energy (USDOE) spent nuclear fuels (SNFs) and their storage/disposal canisters. The technical assessment was based on the thermodynamic properties as well as the chemical and physical characteristics of the materials expected inside the canisters. No chemical reactions were disclosed that could feasibly corrode stainless steel canisters to the point of failure. However, the possibility of embrittlement (loss of ductility) of the stainless steel through contact with liquid metal fission products or hydrogen inside the canisters cannot be dismissed. Higher-than-currently-permitted internal gas pressures must also be considered. These results, based on the assessment of two representative 90-year-cooled fuels that are stored at 200C in stainless steel canisters with internal blankets of helium, may be applied to most of the fuels in the USDOE's SNF inventory

  16. Serbian SNF Repatriation Operation. Issues, Solving, Lesson

    Smirnov, A. [Research and Development Company ' Sosny' , Moscow (Russian Federation)

    2011-07-01

    For now the removal of SNF from RA reactor site (PC NFS, Serbia) is the most time-consuming and technically complicated operation under RRRFR Program. The most efficient techniques and lessons learned from other projects of the RRRFR Program as well as new unique technical decisions were used. Two big challenges were resolved during implementation of Serbian Project: (1) preparation of damaged fuel located in the packages unsuitable for transport, taking into account insufficient infrastructure of RA reactor site and (2) removal of large amount of fuel in one multimodal shipment through several transit countries. The main attention was paid to safety justification of all activities. All approvals were obtained in Russia, Serbia and transit countries. Special canisters were designed for transportation of specific RA reactor fuel (of small dimensions, unidentifiable, damaged due to corrosion). The canister design was selected to be untight - it was the most expedient decision for that case from safety perspective. The technology and a set of equipment were designed for remote removal of the fuel from the existing package (aluminum barrels and reactor channels) and placing of the fuel into the new canisters. After fabrication and assembling of the equipment theoretical and practical training of the personnel was performed. Fuel repackaging took about 5 months. SNF was transported in TUK-19 and SKODA VPVR/M casks. The baskets of large capacity were designed and fabricated for SKODA VPVR/M casks. Special requirements to drying the packages and composition of gaseous medium inside were justified to ensure fire and explosion safety. Specialized ISO-containers and transfer equipment designed under Romanian Project were used together with TUK-19 casks. A forklift and mobile rail system were used to handle SKODA VPVR/M casks under conditions of low capacity of the cranes at the facility. Due to the tight schedule of RRRFR Program as well as geographical peculiarities of RA

  17. A User's Guide to the SNF ampersand INEL EIS

    1995-01-01

    This User's Guide is intended to help you find information in the SNF and INEL EIS (that's short for US Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact Statement). The first section of this Guide gives you a brief overview of the SNF ampersand INEL EIS., The second section is organized to help you find specific information in the Environmental Impact Statement -- whether you're interested in a management alternative, a particular site (such as Hanford), or a discipline (such as land use or water quality)

  18. Nuclear Solid Waste Processing Design at the Idaho Spent Fuels Facility

    Dippre, M. A.

    2003-01-01

    A spent nuclear fuels (SNF) repackaging and storage facility was designed for the Idaho National Engineering and Environmental Laboratory (INEEL), with nuclear solid waste processing capability. Nuclear solid waste included contaminated or potentially contaminated spent fuel containers, associated hardware, machinery parts, light bulbs, tools, PPE, rags, swabs, tarps, weld rod, and HEPA filters. Design of the nuclear solid waste processing facilities included consideration of contractual, regulatory, ALARA (as low as reasonably achievable) exposure, economic, logistical, and space availability requirements. The design also included non-attended transfer methods between the fuel packaging area (FPA) (hot cell) and the waste processing area. A monitoring system was designed for use within the FPA of the facility, to pre-screen the most potentially contaminated fuel canister waste materials, according to contact- or non-contact-handled capability. Fuel canister waste materials which are not able to be contact-handled after attempted decontamination will be processed remotely and packaged within the FPA. Noncontact- handled materials processing includes size-reduction, as required to fit into INEEL permitted containers which will provide sufficient additional shielding to allow contact handling within the waste areas of the facility. The current design, which satisfied all of the requirements, employs mostly simple equipment and requires minimal use of customized components. The waste processing operation also minimizes operator exposure and operator attendance for equipment maintenance. Recently, discussions with the INEEL indicate that large canister waste materials can possibly be shipped to the burial facility without size-reduction. New waste containers would have to be designed to meet the drop tests required for transportation packages. The SNF waste processing facilities could then be highly simplified, resulting in capital equipment cost savings, operational

  19. Current state of WWER SNF storage in Russia and the perspectives

    Anisimov, O.; Kozlov, Y.; Razmashkin, N.; Safutin, V.; Tikhonov, N.

    2006-01-01

    In the Russian Federation WWER-440 Spent Nuclear Fuel (SNF) is reprocessed at RT-1 plant near Cheliabinsk. WWER-1000 SNF is supposed to be reprocessed at RT-2 plant, which will be built about 2020. The information on the capacity and fill up level of the at-reactor pools at NPP with WWER reactors considering its modification up to May 2005 is given. The regulatory requirements to all SNF 'wet' storage facilities; the principle design and engineering solutions as well as the complex of measures for radiation safety and the environmental protection of spent fuel storage are presented. WWER-440 SNF management, WWER-1000 SNF management and dry storage of WWER-1000 SNF are discussed. In the conclusion it is noted than neither Russia, nor any other country have the experience of construction of vault-type 'dry' storage facilities of such a capacity to store WWER-1000 SNF (9000 tU). The experience and design solutions approved earlier in creation of other dangerous facilities were used. The calculations were based on conservative assumptions allowing with a large assurance to guarantee the nuclear and radiation safety and the environmental protection. At present, a program is developed for scientific-technical support of the dry storage facility design and operation, aimed at the studies whose results will allow to optimize the taken technical decisions, simplify SNF management technology and, possibly, to reduce the cost of the storage facility itself

  20. Regulatory practices of radiation safety of SNF transportation in Russia

    Kuryndina, Lidia; Kuryndin, Anton; Stroganov, Anatoly

    2008-01-01

    This paper overviews current regulatory practices for the assurance of nuclear and radiation safety during railway transportation of SNF on the territory of Russian Federation from NPPs to longterm-storage of reprocessing sites. The legal and regulatory requirements (mostly compliant with IAEA ST-1), licensing procedure for NM transportation are discussed. The current procedure does not require a regulatory approval for each particular shipment if the SNF fully comply with the Rosatom's branch standard and is transported in approved casks. It has been demonstrated that SNF packages compliant with the branch standard, which is knowingly provide sufficient safety margin, will conform to the federal level regulations. The regulatory approval is required if a particular shipment does not comply with the branch standard. In this case, the shipment can be approved only after regulatory review of Applicant's documents to demonstrate that the shipment still conformant to the higher level (federal) regulations. The regulatory review frequently needs a full calculation test of the radiation safety assurance. This test can take a lot of time. That's why the special calculation tools were created in SEC NRS. These tools aimed for precision calculation of the radiation safety parameters by SNF transportation use preliminary calculated Green's functions. Such approach allows quickly simulate any source distribution and optimize spent fuel assemblies placement in cask due to the transport equation property of linearity relatively the source. The short description of calculation tools are presented. Also, the paper discusses foreseen implications related to transportation of mixed-oxide SNF. (author)

  1. Dry air oxidation kinetics of K-Basin spent nuclear fuel

    Abrefah, J.; Buchanan, H.C.; Gerry, W.M.; Gray, W.J.; Marschman, S.C.

    1998-06-01

    The safety and process analyses of the proposed Integrated Process Strategy (IPS) to move the N-Reactor spent nuclear fuel (SNF) stored at K-Basin to an interim storage facility require information about the oxidation behavior of the metallic uranium. Limited experiments have been performed on the oxidation reaction of SNF samples taken from an N-Reactor outer fuel element in various atmospheres. This report discusses studies on the oxidation behavior of SNF using two independent experimental systems: (1) a tube furnace with a flowing gas mixture of 2% oxygen/98% argon; and (2) a thermogravimetric system for dry air oxidation

  2. DOE-owned spent nuclear fuel program plan

    1995-11-01

    The Department of Energy (DOE) has produced spent nuclear fuel (SNF) for many years as part of its various missions and programs. The historical process for managing this SNF was to reprocess it whereby valuable material such as uranium or plutonium was chemically separated from the wastes. These fuels were not intended for long-term storage. As the need for uranium and plutonium decreased, it became necessary to store the SNF for extended lengths of time. This necessity resulted from a 1992 DOE decision to discontinue reprocessing SNF to recover strategic materials (although limited processing of SNF to meet repository acceptance criteria remains under consideration, no plutonium or uranium extraction for other uses is planned). Both the facilities used for storage, and the fuel itself, began experiencing aging from this extended storage. New efforts are now necessary to assure suitable fuel and facility management until long-term decisions for spent fuel disposition are made and implemented. The Program Plan consists of 14 sections as follows: Sections 2--6 describe objectives, management, the work plan, the work breakdown structure, and the responsibility assignment matrix. Sections 7--9 describe the program summary schedules, site logic diagram, SNF Program resource and support requirements. Sections 10--14 present various supplemental management requirements and quality assurance guidelines

  3. DOE-owned spent nuclear fuel program plan

    NONE

    1995-11-01

    The Department of Energy (DOE) has produced spent nuclear fuel (SNF) for many years as part of its various missions and programs. The historical process for managing this SNF was to reprocess it whereby valuable material such as uranium or plutonium was chemically separated from the wastes. These fuels were not intended for long-term storage. As the need for uranium and plutonium decreased, it became necessary to store the SNF for extended lengths of time. This necessity resulted from a 1992 DOE decision to discontinue reprocessing SNF to recover strategic materials (although limited processing of SNF to meet repository acceptance criteria remains under consideration, no plutonium or uranium extraction for other uses is planned). Both the facilities used for storage, and the fuel itself, began experiencing aging from this extended storage. New efforts are now necessary to assure suitable fuel and facility management until long-term decisions for spent fuel disposition are made and implemented. The Program Plan consists of 14 sections as follows: Sections 2--6 describe objectives, management, the work plan, the work breakdown structure, and the responsibility assignment matrix. Sections 7--9 describe the program summary schedules, site logic diagram, SNF Program resource and support requirements. Sections 10--14 present various supplemental management requirements and quality assurance guidelines.

  4. Ultimate disposition of aluminum clad spent nuclear fuel in the United States

    Messick, C.E.; Clark, W.D.; Clapper, M.; Mustin, T.P.

    2001-01-01

    Treatment and disposition of spent nuclear fuel (SNF) in the United States has changed significantly over the last decade due to change in world climate associated with nuclear material. Chemical processing of aluminum based SNF is ending and alternate disposition paths are being developed that will allow for the ultimate disposition of the enriched uranium in this SNF. Existing inventories of aluminum based SNF are currently being stored primarily in water-filled basins at the Savannah River Site (SRS) while these alternate disposition paths are being developed and implemented. Nuclear nonproliferation continues to be a worldwide concern and it is causing a significant influence on the development of management alternatives for SNF. SRS recently completed an environmental impact statement for the management of aluminum clad SNF that selects alternatives for all of the fuels in inventory. The U.S. Department of Energy and SRS are now implementing a dual strategy of processing small quantities of 'problematic' SNF while developing an alternative technology to dispose of the remaining aluminum clad SNF in the proposed monitored geologic repository. (author)

  5. A secondary fuel removal process: plasma processing

    Min, J Y; Kim, Y S [Hanyang Univ., Seoul (Korea, Republic of); Bae, K K; Yang, M S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    Plasma etching process of UO{sub 2} by using fluorine containing gas plasma is studied as a secondary fuel removal process for DUPIC (Direct Use of PWR spent fuel Into Candu) process which is taken into consideration for potential future fuel cycle in Korea. CF{sub 4}/O{sub 2} gas mixture is chosen for reactant gas and the etching rates of UO{sub 2} by the gas plasma are investigated as functions of CF{sub 4}/O{sub 2} ratio, plasma power, substrate temperature, and plasma gas pressure. It is found that the optimum CF{sub 4}/O{sub 2} ratio is around 4:1 at all temperatures up to 400 deg C and the etching rate increases with increasing r.f. power and substrate temperature. Under 150W r.f. power the etching rate reaches 1100 monolayers/min at 400 deg C, which is equivalent to about 0.5mm/min. (author).

  6. SNF project's MCO compliance assessment with DOE ''general design criteria,'' order 6430.1A and ''SNF project MCO additional NRC requirements,'' HNF-SD-SNF-DB-005

    GOLDMANN, L.H.

    1999-01-01

    This document is presented to demonstrate the MCOs compliance to the major design criteria invoked on the MCO. This document is broken down into a section for the MCO's evaluation against DOE Order 6430.1A General Design Criteria sixteen divisions and then the evaluation of the MCO against HNF-SD-SNF-DB-005 ''Spent Nuclear Fuel Project Multi-Canister Overpack Additional NRC Requirements.'' The compliance assessment is presented as a matrix in tabular form. The MCO is the primary container for the K-basin's spent nuclear fuel as it leaves the basin pools and through to the 40 year interim storage at the Canister Storage Building (CSB). The MCO and its components interface with; the K basins, shipping cask and transportation system, Cold Vacuum Drying facility individual process bays and equipment, and CSB facility including the MCO handling machine (MHM), the storage tubes, and the MCO work stations where sampling, welding, and inspection of the MCO is performed. As the MCO is the primary boundary for handling, process, and storage, its main goals are to minimize the spread of its radiological contents to the outside of the MCO and provide for nuclear criticality control. The MCO contains personnel radiation shielding only on its upper end, in the form of a shield plug, where the process interfaces are located. Shielding beyond the shield plug is the responsibility of the using facilities. The design of the MCO and its components is depicted in drawings H-2-828040 through H-2-828075. Not every drawing number in the sequence is used. The first drawing number, H-2-828040, is the drawing index for the MCO. The design performance specification for the MCO is HW-S-0426, and was reviewed and approved by the interfacing design authorities, the safety, regulatory, and operations groups, and the local DOE office. The current revision for the design performance specification is revision 5. The designs of the MCO have been reviewed and approved in a similar way and the reports

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

    Hill, Thomas J

    2005-09-01

    interim dry storage facilities, thus permitting the shutdown and decommission of the older facilities. Two wet pool facilities, one at the Idaho Nuclear Technology and Engineering Center and the other at Test Area North, were targeted for initial SNF movements since these were some of the oldest at the INL. Because of the difference in the SNF materials different types of drying processes had to be developed. Passive drying, as is done with typical commercial SNF was not an option because on the condition of some of the fuel, the materials to be dried, and the low heat generation of some of the SNF. There were also size limitations in the existing facility. Active dry stations were designed to address the specific needs of the SNF and the facilities.

  8. Spent Nuclear Fuel Project Safety Management Plan

    Garvin, L.J.

    1996-02-01

    The Spent Nuclear Fuel Project Safety Management Plan describes the new nuclear facility regulatory requirements basis for the Spemt Nuclear Fuel (SNF) Project and establishes the plan to achieve compliance with this basis at the new SNF Project facilities

  9. Main Principles of the Perspective System of SNF Management in Russia - 13333

    Baryshnikov, Mikhail

    2013-01-01

    For the last several years the System of the Spent Nuclear Fuel management in Russia was seriously changed. The paper describes the main principles of the changes and the bases of the Perspective System of SNF Management in Russia. Among such the bases there are the theses with the interesting names like 'total knowledge', 'pollutant pays' and 'pay and forget'. There is also a brief description of the modern Russian SNF Management Infrastructure. And an outline of the whole System. The System which is - in case of Russia - is quite necessary to adjust SNF accumulation and to utilize the nuclear heritage. (authors)

  10. HANSF 1.3 Users Manual FAI/98-40-R2 Hanford Spent Nuclear Fuel (SNF) Safety Analysis Model [SEC 1 and 2

    DUNCAN, D.R.

    1999-10-07

    The HANSF analysis tool is an integrated model considering phenomena inside a multi-canister overpack (MCO) spent nuclear fuel container such as fuel oxidation, convective and radiative heat transfer, and the potential for fission product release. This manual reflects the HANSF version 1.3.2, a revised version of 1.3.1. HANSF 1.3.2 was written to correct minor errors and to allow modeling of condensate flow on the MCO inner surface. HANSF 1.3.2 is intended for use on personal computers such as IBM-compatible machines with Intel processors running under Lahey TI or digital Visual FORTRAN, Version 6.0, but this does not preclude operation in other environments.

  11. Legal precedents regarding use and defensibility of risk assessment in Federal transportation of SNF and HLW

    Bentz, E.J. Jr.; Bentz, C.B.; O'Hora, T.D.; Chen, S.Y.

    1997-01-01

    Risk assessment has become an increasingly important and essential tool in support of Federal decision-making regarding the handling, storage, disposal, and transportation of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). This paper analyzes the current statutory and regulatory framework and related legal precedents with regard to SNF and HLW transportation. The authors identify key scientific and technical issues regarding the use and defensibility of risk assessment in Federal decision-making regarding anticipated shipments

  12. Enhanced amino acid utilization sustains growth of cells lacking Snf1/AMPK

    Nicastro, Raffaele; Tripodi, Farida; Guzzi, Cinzia

    2015-01-01

    when grown with glucose excess. We show that loss of Snf1 in cells growing in 2% glucose induces an extensive transcriptional reprogramming, enhances glycolytic activity, fatty acid accumulation and reliance on amino acid utilization for growth. Strikingly, we demonstrate that Snf1/AMPK-deficient cells...... remodel their metabolism fueling mitochondria and show glucose and amino acids addiction, a typical hallmark of cancer cells....

  13. Trehalose-6-phosphate synthesis controls yeast gluconeogenesis downstream and independent of SNF1.

    Deroover, Sofie; Ghillebert, Ruben; Broeckx, Tom; Winderickx, Joris; Rolland, Filip

    2016-06-01

    Trehalose-6-P (T6P), an intermediate of trehalose biosynthesis, was identified as an important regulator of yeast sugar metabolism and signaling. tps1Δ mutants, deficient in T6P synthesis (TPS), are unable to grow on rapidly fermentable medium with uncontrolled influx in glycolysis, depletion of ATP and accumulation of sugar phosphates. However, the exact molecular mechanisms involved are not fully understood. We show that SNF1 deletion restores the tps1Δ growth defect on glucose, suggesting that lack of TPS hampers inactivation of SNF1 or SNF1-regulated processes. In addition to alternative, non-fermentable carbon metabolism, SNF1 controls two major processes: respiration and gluconeogenesis. The tps1Δ defect appears to be specifically associated with deficient inhibition of gluconeogenesis, indicating more downstream effects. Consistently, Snf1 dephosphorylation and inactivation on glucose medium are not affected, as confirmed with an in vivo Snf1 activity reporter. Detailed analysis shows that gluconeogenic Pck1 and Fbp1 expression, protein levels and activity are not repressed upon glucose addition to tps1Δ cells, suggesting a link between the metabolic defect and persistent gluconeogenesis. While SNF1 is essential for induction of gluconeogenesis, T6P/TPS is required for inactivation of gluconeogenesis in the presence of glucose, downstream and independent of SNF1 activity and the Cat8 and Sip4 transcription factors. © FEMS 2016. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

  14. Dry Process Fuel Performance Evaluation

    Yang, Myung Seung; Song, K. C.; Moon, J. S. and others

    2005-04-15

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  15. Dry Process Fuel Performance Evaluation

    Yang, Myung Seung; Song, K. C.; Moon, J. S. and others

    2005-04-01

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  16. Collective processing device for spent fuel

    Irie, Hiroaki; Taniguchi, Noboru.

    1996-01-01

    The device of the present invention comprises a sealing vessel, a transporting device for transporting spent fuels to the sealing vessel, a laser beam cutting device for cutting the transported spent fuels, a dissolving device for dissolving the cut spent fuels, and a recovering device for recovering radioactive materials from the spent fuels during processing. Reprocessing treatments comprising each processing of dismantling, shearing and dissolving are conducted in the sealing vessel can ensure a sealing barrier for the radioactive materials (fissionable products and heavy nuclides). Then, since spent fuels can be processed in a state of assemblies, and the spent fuels are easily placed in the sealing vessel, operation efficiency is improved, as well as operation cost is saved. Further, since the spent fuels can be cut by a remote laser beam operation, there can be prevented operator's exposure due to radioactive materials released from the spent fuels during cutting operation. (T.M.)

  17. Review of technologies and alternatives for dispositioning DOE/INEL spent nuclear fuels

    Bendixsen, C.L.; Olson, A.L.; Christian, J.D.; Thomas, T.R.

    1994-01-01

    Following the decision to cease processing of spent nuclear fuels (SNF) for purposes of uranium recovery, the Idaho National Engineering Laboratory is evaluating alternatives for management of the 90+ types of SNF now stored thereon. Two major classes of alternatives include direct disposal of SNF with little or minimal repackaging and chemical processing to reduce high-activity waste volumes and to reduce the number and type of waste forms requiring regulatory approval. The disposal alternatives are described and the primary advantages and disadvantages identified

  18. Dry process fuel performance technology development

    Kang, Kweon Ho; Kim, K. W.; Kim, B. K. (and others)

    2006-06-15

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

  19. Dry process fuel performance technology development

    Kang, Kweon Ho; Kim, K. W.; Kim, B. K.

    2006-06-01

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

  20. Spent fuel storage process equipment development

    Park, Hyun Soo; Lee, Jae Sol; Yoo, Jae Hyung

    1990-02-01

    Nuclear energy which is a major energy source of national energy supply entails spent fuels. Spent fuels which are high level radioactive meterials, are tricky to manage and need high technology. The objectives of this study are to establish and develop key elements of spent fuel management technologies: handling equipment and maintenance, process automation technology, colling system, and cleanup system. (author)

  1. Assessment results of the Indonesian TRIGA SNF to be shipped to INEEL

    Jefimoff, J.; Robb, A.K.; Wendt, K.M.; Syarip, I.; Alfa, T.

    1997-01-01

    This paper describes the Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) examination performed by technical personnel from the Idaho National Engineering and Environmental Laboratory (INEEL) at the Bandung and Yogyakarta research reactor facilities in Indonesia. The examination was required before the SNF would be accepted for transportation to and storage at the INEEL. This paper delineates the Initial Preparations prior to the Indonesian foreign research reactor (FRR) fuel examination. The technical basis for the examination, the TRIGA SNF Acceptance Criteria, and the physical condition required for transportation, receipt and storage of the TRIGA SNF at the INEEL is explained. In addition to the initial preparations, preparation descriptions of the Work Plan For TRIGA Fuel Examination, the Underwater Examination Equipment used, and personnel Examination Team Training are included. Finally, the Fuel Examination and Results of the aluminum and stainless steel clad TRIGA fuel examination have been summarized. Lessons learned from all the activities completed to date is provided in an addendum. The initial preparations included: (1) coordination between the INEEL, FRR or Badan Tenaga Atom Nasional (BATAN), DOE-HQ, and the US State Department and Embassy; (2) incorporating Savannah River Site (SRS) FRR experience and lessons learned; (3) collecting both FRR facility and spent fuel data, and issuing a radionuclide report (Radionuclide Mass Inventory, Activity, Decay Heat, and Dose Rate Parametric Data for TRIGA Spent Nuclear Fuels) needed for transportation and fuel acceptance at the INEEL; and (4) preexamination work at the research reactor for the fuel examination

  2. Microstructure and Physical Metallurgy in Melt-Dilute Treatment Technology for Al-Based Spent Nuclear Fuel

    Adams, T.M.; Peacock, H.B.

    1998-09-01

    Major challenges associated with the direct disposal of research reactor fuel in a repository include nonproliferation and criticality control, both of which may be a concern for HEU Al-SNF. Consideration must be given to the potential desirability and/or regulatory necessity of diluting the HEU SNF to below 20 percent enrichment. The probability for a criticality event as well as the issue of proliferation are greatly decreased by reducing the enrichment. The melt-dilute technology development program is focused on the development and implementation of a treatment technology for diluting HEU Al-SNF to LEU levels and qualifying this LEU Al-SNF form for geologic repository storage. An attribute of the melt-dilute process is its applicability to various types of FRR and DRR SNF

  3. Technical, economical and legal aspects of repatriation of Russian-origin research reactor SNF to Russia

    Smirnov, A.; Kanashov, B.; Efarov, S.; Lebedev, A.; Kolupaev, D.

    2005-01-01

    The aim of the report is to find some principal decisions to implement an Agreement between the Governments of the Russian Federation and the USA on repatriation of the research reactor spent nuclear fuel (RR SNF) to the Russian Federation. The report presents some ideas and approaches to the transportation of the Russian-origin RR SNF from the technical, economical and legal viewpoints. The report summarizes the Russian experience and possibilities to fulfill the program under the Agreement. Some decisions are proposed related to application of the international transportation experience and the most advanced technologies for the RR SNF handling. At present, there is no any unified SNF transportation technology that is capable to implement the transportation program schedule set by the Agreement. The decision is in the comprehensive approach as well as in the development of mobile and flexible schemes and in implementation of parallel and combined shipments. (author)

  4. A Green Approach to SNF Reprocessing: Are Common Household Reagents the Answer?

    Peper, Shane M.; McNamara, Bruce K.; O'Hara, Matthew J.; Douglas, Matthew

    2008-01-01

    It has been discovered that UO2, the principal component of spent nuclear fuel (SNF), can efficiently be dissolved at room temperature using a combination of common household reagents, namely hydrogen peroxide, baking soda, and ammonia. This rather serendipitous discovery opens up the possibility, for the first time, of considering a non-acidic process for recycling U from SNF. Albeit at the early stages of development, our unconventional dissolution approach possesses many attractive features that could make it a reality in the future. With dissolution byproducts of water and oxygen, our approach poses a minimal threat to the environment. Moreover, the use of common household reagents to afford actinide oxide dissolution suggests a certain degree of economic favorability. With the use of a ''closed'' digestion vessel as a reaction chamber, our approach has substantial versatility with the option of using either aqueous or gaseous reactant feeds or a combination of both. Our approach distinguishes itself from all existing reprocessing technologies in two important ways. First and foremost, it is an alkaline rather than an acidic process, using mild non-corrosive chemicals under ambient conditions to effect actinide separations. Secondly, it does not dissolve the entire SNF matrix, but rather selectively solubilizes U and other light actinides for subsequent separation, resulting in potentially faster head-end dissolution and fewer downstream separation steps. From a safeguards perspective, the use of oxidizing alkaline solutions to effect actinide separations also potentially offers a degree of inherent proliferation resistance, by allowing the U to be selectively removed from the remaining dissolver solution while keeping Pu grouped with the other minor actinides and fission products. This paper will describe the design and general experimental setup of a 'closed' digestion vessel for performing uranium oxide dissolutions under alkaline conditions using gaseous

  5. A Green Approach to SNF Reprocessing: Are Common Household Reagents the Answer?

    Peper, Shane M.; McNamara, Bruce K.; O' Hara, Matthew J.; Douglas, Matthew

    2008-04-03

    It has been discovered that UO2, the principal component of spent nuclear fuel (SNF), can efficiently be dissolved at room temperature using a combination of common household reagents, namely hydrogen peroxide, baking soda, and ammonia. This rather serendipitous discovery opens up the possibility, for the first time, of considering a non-acidic process for recycling U from SNF. Albeit at the early stages of development, our unconventional dissolution approach possesses many attractive features that could make it a reality in the future. With dissolution byproducts of water and oxygen, our approach poses a minimal threat to the environment. Moreover, the use of common household reagents to afford actinide oxide dissolution suggests a certain degree of economic favorability. With the use of a “closed” digestion vessel as a reaction chamber, our approach has substantial versatility with the option of using either aqueous or gaseous reactant feeds or a combination of both. Our approach distinguishes itself from all existing reprocessing technologies in two important ways. First and foremost, it is an alkaline rather than an acidic process, using mild non-corrosive chemicals under ambient conditions to effect actinide separations. Secondly, it does not dissolve the entire SNF matrix, but rather selectively solubilizes U and other light actinides for subsequent separation, resulting in potentially faster head-end dissolution and fewer downstream separation steps. From a safeguards perspective, the use of oxidizing alkaline solutions to effect actinide separations also potentially offers a degree of inherent proliferation resistance, by allowing the U to be selectively removed from the remaining dissolver solution while keeping Pu grouped with the other minor actinides and fission products. This paper will describe the design and general experimental setup of a “closed” digestion vessel for performing uranium oxide dissolutions under alkaline conditions using

  6. Fuel corrosion processes under waste disposal conditions

    Shoesmith, D.W.

    2000-01-01

    The release of the majority of radionuclides from spent nuclear fuel under permanent disposal conditions will be controlled by the rate of dissolution of the UO 2 fuel matrix. In this manuscript the mechanism of the coupled anodic (fuel dissolution) and cathodic (oxidant reduction) reactions which constitute the overall fuel corrosion process is reviewed, and the many published observations on fuel corrosion under disposal conditions discussed. The primary emphasis is on summarizing the overall mechanistic behaviour and establishing the primary factors likely to control fuel corrosion. Included are discussions on the influence of various oxidants including radiolytic ones, pH, temperature, groundwater composition, and the formation of corrosion product deposits. The relevance of the data recorded on unirradiated UO 2 to the interpretation of spent fuel behaviour is included. Based on the review, the data used to develop fuel corrosion models under the conditions anticipated in Yucca Mountain (NV, USA) are evaluated

  7. Fuel quality processing study, volume 1

    Ohara, J. B.; Bela, A.; Jentz, N. E.; Syverson, H. T.; Klumpe, H. W.; Kessler, R. E.; Kotzot, H. T.; Loran, B. L.

    1981-01-01

    A fuel quality processing study to provide a data base for an intelligent tradeoff between advanced turbine technology and liquid fuel quality, and also, to guide the development of specifications of future synthetic fuels anticipated for use in the time period 1985 to 2000 is given. Four technical performance tests are discussed: on-site pretreating, existing refineries to upgrade fuels, new refineries to upgrade fuels, and data evaluation. The base case refinery is a modern Midwest refinery processing 200,000 BPD of a 60/40 domestic/import petroleum crude mix. The synthetic crudes used for upgrading to marketable products and turbine fuel are shale oil and coal liquids. Of these syncrudes, 50,000 BPD are processed in the existing petroleum refinery, requiring additional process units and reducing petroleum feed, and in a new refinery designed for processing each syncrude to produce gasoline, distillate fuels, resid fuels, and turbine fuel, JPGs and coke. An extensive collection of synfuel properties and upgrading data was prepared for the application of a linear program model to investigate the most economical production slate meeting petroleum product specifications and turbine fuels of various quality grades. Technical and economic projections were developed for 36 scenarios, based on 4 different crude feeds to either modified existing or new refineries operated in 2 different modes to produce 7 differing grades of turbine fuels. A required product selling price of turbine fuel for each processing route was calculated. Procedures and projected economics were developed for on-site treatment of turbine fuel to meet limitations of impurities and emission of pollutants.

  8. Implementation of Agile project management in spent nuclear fuel characterization process

    Vinas Pena, P.

    2015-01-01

    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

  9. Manufacturing technology and process for BWR fuel

    Kato, Shigeru

    1996-01-01

    Following recent advanced technologies, processes and requests of the design changes of BWR fuel, Nuclear Fuel Industries, Ltd. (NFI) has upgraded the manufacturing technology and honed its own skills to complete its brand-new automated facility in Tokai in the latter half of 1980's. The plant uses various forms of automation throughout the manufacturing process: the acceptance of uranium dioxide powder, pelletizing, fuel rod assembling, fuel bundle assembling and shipment. All processes are well computerized and linked together to establish the integrated control system with three levels of Production and Quality Control, Process Control and Process Automation. This multi-level system plays an important role in the quality assurance system which generates the highest quality of fuels and other benefits. (author)

  10. Fuel processing requirements and techniques for fuel cell propulsion power

    Kumar, R.; Ahmed, S.; Yu, M.

    Fuels for fuel cells in transportation systems are likely to be methanol, natural gas, hydrogen, propane, or ethanol. Fuels other than hydrogen will need to be reformed to hydrogen on-board the vehicle. The fuel reformer must meet stringent requirements for weight and volume, product quality, and transient operation. It must be compact and lightweight, must produce low levels of CO and other byproducts, and must have rapid start-up and good dynamic response. Catalytic steam reforming, catalytic or noncatalytic partial oxidation reforming, or some combination of these processes may be used. This paper discusses salient features of the different kinds of reformers and describes the catalysts and processes being examined for the oxidation reforming of methanol and the steam reforming of ethanol. Effective catalysts and reaction conditions for the former have been identified; promising catalysts and reaction conditions for the latter are being investigated.

  11. Status of burnup credit for transport of SNF in the United States

    Parks, C.V.; Wagner, J.C.

    2004-01-01

    Allowing credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transportation, and disposal of spent nuclear fuel (SNF) while maintaining a subcritical margin sufficient to establish an adequate safety basis. This paper reviews the current status of burnup credit applied to the design and transport of SNF casks in the United States. The existing U.S. regulatory guidance on burnup credit is limited to pressurized-water-reactor (PWR) fuel and to allowing credit only for actinides in the SNF. By comparing loading curves against actual SNF discharge data for U.S. reactors, the potential benefits that can be realized using the current regulatory guidance with actinide-only burnup credit are illustrated in terms of the inventory allowed in high-capacity casks and the concurrent reduction in SNF shipments. The additional benefits that might be realized by extending burnup credit to credit for select fission products are also illustrated. The curves show that, although fission products in SNF provide a small decrease in reactivity compared with actinides, the additional negative reactivity causes the SNF inventory acceptable for transportation to increase from roughly 30% to approximately 90% when fission products are considered. A savings of approximately $150M in transport costs can potentially be realized for the planned inventory of the repository. Given appropriate experimental data to support code validation, a realistic best-estimate analysis of burnup credit that includes validated credit for fission products is the enhancement that will yield the most significant impact on future transportation plans

  12. Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements

    KLEM, M.J.

    2000-01-01

    In 1998, a major change in the technical strategy for managing Multi Canister Overpacks (MCO) while stored within the Canister Storage Building (CSB) occurred. The technical strategy is documented in Baseline Change Request (BCR) No. SNF-98-006, Simplified SNF Project Baseline (MCO Sealing) (FDH 1998). This BCR deleted the hot conditioning process initially adopted for the Spent Nuclear Fuel Project (SNF Project) as documented in WHC-SD-SNF-SP-005, Integrated Process Strategy for K Basins Spent Nuclear Fuel (WHC 199.5). In summary, MCOs containing Spent Nuclear Fuel (SNF) from K Basins would be placed in interim storage following processing through the Cold Vacuum Drying (CVD) facility. With this change, the needs for the Hot Conditioning System (HCS) and inerting/pressure retaining capabilities of the CSB storage tubes and the MCO Handling Machine (MHM) were eliminated. Mechanical seals will be used on the MCOs prior to transport to the CSB. Covers will be welded on the MCOs for the final seal at the CSB. Approval of BCR No. SNF-98-006, imposed the need to review and update the CSB functions and requirements baseline documented herein including changing the document title to ''Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements.'' This revision aligns the functions and requirements baseline with the CSB Simplified SNF Project Baseline (MCO Sealing). This document represents the Canister Storage Building (CSB) Subproject technical baseline. It establishes the functions and requirements baseline for the implementation of the CSB Subproject. The document is organized in eight sections. Sections 1.0 Introduction and 2.0 Overview provide brief introductions to the document and the CSB Subproject. Sections 3.0 Functions, 4.0 Requirements, 5.0 Architecture, and 6.0 Interfaces provide the data described by their titles. Section 7.0 Glossary lists the acronyms and defines the terms used in this document. Section 8.0 References lists the

  13. Powder handling for automated fuel processing

    Frederickson, J.R.; Eschenbaum, R.C.; Goldmann, L.H.

    1989-01-01

    Installation of the Secure Automated Fabrication (SAF) line has been completed. It is located in the Fuel Cycle Plant (FCP) at the Department of Energy's (DOE) Hanford site near Richland, Washington. The SAF line was designed to fabricate advanced reactor fuel pellets and assemble fuel pins by automated, remote operation. This paper describes powder handling equipment and techniques utilized for automated powder processing and powder conditioning systems in this line. 9 figs

  14. Complexon Solutions in Freon for Decontamination of Solids and SNF Treatment

    Kamachev, V.; Shadrin, A.; Murzin, A.

    2008-01-01

    Full text of publication follows: The possibility of using complexon solutions in supercritical and compressed carbon dioxide for decontamination of solid surfaces and for spent nuclear fuel (SNF) treatment was demonstrated in the works of Japanese, Russian and American researchers. The obtained data showed that the use of complexon solutions in carbon dioxide sharply decreases the volume of secondary radioactive wastes because it can be easily evaporated, purified and recycled. Moreover, high penetrability of carbon dioxide allows decontamination of surfaces with complex shape. However, one of the disadvantages of carbon dioxide is its high working pressure (10-20 MPa for supercritical CO 2 and 7 MPa for compressed CO 2 ). Moreover, in case of SNF treatment, carbon dioxide solvent will be contaminated with 14 C, which in the course of SNF dissolution in CO 2 containing TBP*HNO 3 adduct stage will be oxidized into CO 2 . These main disadvantages can be eliminated by using complexon solutions in ozone-friendly Freon HFC-134a for decontamination and SNF treatment. Our experimental data for real contaminated materials showed that the decontamination factor for complexon solutions in liquid Freon HFC-134a at 1,2 MPa and 25 deg. C is close to that attained in carbon dioxide. Moreover, the possibility of SNF treatment in Freon HFC-134a was demonstrated in trials using real SNF and its imitators. (authors)

  15. Mathematical modeling of biomass fuels formation process

    Gaska, Krzysztof; Wandrasz, Andrzej J.

    2008-01-01

    The increasing demand for thermal and electric energy in many branches of industry and municipal management accounts for a drastic diminishing of natural resources (fossil fuels). Meanwhile, in numerous technical processes, a huge mass of wastes is produced. A segregated and converted combustible fraction of the wastes, with relatively high calorific value, may be used as a component of formed fuels. The utilization of the formed fuel components from segregated groups of waste in associated processes of co-combustion with conventional fuels causes significant savings resulting from partial replacement of fossil fuels, and reduction of environmental pollution resulting directly from the limitation of waste migration to the environment (soil, atmospheric air, surface and underground water). The realization of technological processes with the utilization of formed fuel in associated thermal systems should be qualified by technical criteria, which means that elementary processes as well as factors of sustainable development, from a global viewpoint, must not be disturbed. The utilization of post-process waste should be preceded by detailed technical, ecological and economic analyses. In order to optimize the mixing process of fuel components, a mathematical model of the forming process was created. The model is defined as a group of data structures which uniquely identify a real process and conversion of this data in algorithms based on a problem of linear programming. The paper also presents the optimization of parameters in the process of forming fuels using a modified simplex algorithm with a polynomial worktime. This model is a datum-point in the numerical modeling of real processes, allowing a precise determination of the optimal elementary composition of formed fuels components, with assumed constraints and decision variables of the task

  16. Microwave processing in MOX fuel cycle

    Mallik, G.K.; Malav, R.K.; Panakkal, J.P.; Kamath, H.S.

    2005-01-01

    The prominent aspect of the microwave heating technique applications in nuclear material processing is its eco-friendly status. It is envisaged that no active liquid waste will be generated from microwave processing. AFFF has fabricated the (U, Pu) 2 O mixed oxide fuels for PHWRs, BWRs and PFBR. AFFF is also working for the AHWR fuel cycle. The present paper summarises about the process experiments, instrumental development, results, and future applications of microwave heating technique. (author)

  17. Chemical aspects of nuclear fuel fabrication processes

    Naylor, A; Ellis, J F; Watson, R H

    1986-04-01

    Processes used by British Nuclear Fuels plc for the conversion of uranium ore concentrates to uranium metal and uranium hexafluoride, are reviewed. Means of converting the latter compound, after enrichment, to sintered UO/sub 2/ fuel bodies are also described. An overview is given of the associated chemical engineering technology.

  18. Fuel corrosion processes under waste disposal conditions

    Shoesmith, D.W.

    1999-09-01

    Under the oxidizing conditions likely to be encountered in the Yucca Mountain Repository, fuel dissolution is a corrosion process involving the coupling of the anodic dissolution of the fuel with the cathodic reduction of oxidants available within the repository. The oxidants potentially available to drive fuel corrosion are environmental oxygen, supplied by the transport through the permeable rock of the mountain and molecular and radical species produced by the radiolysis of available aerated water. The mechanism of these coupled anodic and cathodic reactions is reviewed in detail. While gaps in understanding remain, many kinetic features of these reactions have been studied in considerable detail, and a reasonably justified mechanism for fuel corrosion is available. The corrosion rate is determined primarily by environmental factors rather than the properties of the fuel. Thus, with the exception of increase in rate due to an increase in surface area, pre-oxidation of the fuel has little effect on the corrosion rate

  19. Fuel corrosion processes under waste disposal conditions

    Shoesmith, D.W. [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada)

    1999-09-01

    Under the oxidizing conditions likely to be encountered in the Yucca Mountain Repository, fuel dissolution is a corrosion process involving the coupling of the anodic dissolution of the fuel with the cathodic reduction of oxidants available within the repository. The oxidants potentially available to drive fuel corrosion are environmental oxygen, supplied by the transport through the permeable rock of the mountain and molecular and radical species produced by the radiolysis of available aerated water. The mechanism of these coupled anodic and cathodic reactions is reviewed in detail. While gaps in understanding remain, many kinetic features of these reactions have been studied in considerable detail, and a reasonably justified mechanism for fuel corrosion is available. The corrosion rate is determined primarily by environmental factors rather than the properties of the fuel. Thus, with the exception of increase in rate due to an increase in surface area, pre-oxidation of the fuel has little effect on the corrosion rate.

  20. SWI/SNF complex in disorder

    Santen, Gijs W.E.; Kriek, Marjolein; van Attikum, Haico

    2012-01-01

    Heterozygous germline mutations in components of switch/sucrose nonfermenting (SWI/SNF) chromatin remodeling complexes were recently identified in patients with non-syndromic intellectual disability, Coffin-Siris syndrome and Nicolaides-Baraitser syndrome. The common denominator of the phenotype of these patients is severe intellectual disability and speech delay. Somatic and germline mutations in SWI/SNF components were previously implicated in tumor development. This raises the question whether patients with intellectual disability caused by SWI/SNF mutations in the germline are exposed to an increased risk of developing cancer. Here we compare the mutational spectrum of SWI/SNF components in intellectual disability syndromes and cancer, and discuss the implications of the results of this comparison for the patients. PMID:23010866

  1. Fuel processing for molten-salt reactors

    Hightower, J.R. Jr.

    1976-01-01

    Research devoted to development of processes for the isolation of protactinium and for the removal of fission products from molten-salt breeder reactors is reported. During this report period, engineering development progressed on continuous fluorinators for uranium removal, the metal transfer process for rare-earth removal, the fuel reconstitution step, and molten salt--bismuth contactors to be used in reductive extraction processes. The metal transfer experiment MTE-3B was started. In this experiment all parts of the metal transfer process for rare-earth removal are demonstrated using salt flow rates which are about 1 percent of those required to process the fuel salt in a 1000-MW(e) MSBR. During this report period the salt and bismuth phases were transferred to the experimental vessels, and two runs with agitator speeds of 5 rps were made to measure the rate of transfer of neodymium from the fluoride salt to the Bi--Li stripper solution. The uranium removed from the fuel salt by fluorination must be returned to the processed salt in the fuel reconstitution step before the fuel salt is returned to the reactor. An engineering experiment to demonstrate the fuel reconstitution step is being installed. In this experiment gold-lined equipment will be used to avoid introducing products of corrosion by UF 6 and UF 5 . Alternative methods for providing the gold lining include electroplating and mechanical fabrication

  2. Process for producing nuclear reactor fuel oxides

    Goenrich, H.; Druckenbrodt, W.G.

    1981-01-01

    The waste gases of the calcination process furnace in the AVC or AV/PuC process (manufacture of nuclear reactor fuel dioxides) are returned to the furnace in a closed circuit. The NH 3 produced replaces the hydrogen which would otherwise be required for reduction in this process. (orig.) [de

  3. FY17 Status Report: Research on Stress Corrosion Cracking of SNF Interim Storage Canisters.

    Schindelholz, Eric John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alexander, Christopher L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-01

    This progress report describes work done in FY17 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. Work in FY17 refined our understanding of the chemical and physical environment on canister surfaces, and evaluated the relationship between chemical and physical environment and the form and extent of corrosion that occurs. The SNL corrosion work focused predominantly on pitting corrosion, a necessary precursor for SCC, and process of pit-to-crack transition; it has been carried out in collaboration with university partners. SNL is collaborating with several university partners to investigate SCC crack growth experimentally, providing guidance for design and interpretation of experiments.

  4. Probabilistic Risk Assessment of Cask Drop Accident during On-site Spent Nuclear Fuel Transportation

    Ham, Jae Hyun; Christian, Robby; Momani, Belal Al; Kang, Hyun Gook [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    There are two ways to transfer the SNF from a site to other site, one is land transportation and the other is maritime transportation. Maritime transportation might be used because this way uses more safe route which is far from populated area. The whole transportation process can be divided in two parts: transferring the SNF between SNP and wharf in-Nuclear Power Plant (NPP) site by truck, and transferring the SNF from the wharf to the other wharf by ship. In this research, on-site SNF transportation between SNP and wharf was considered. Two kinds of single accident can occur during this type of SNF transportation, impact and fire, caused by internal events and external events. In this research, PRA of cask drop accident during onsite SNF transportation was done, risk to a person (mSv/person) from a case with specific conditions was calculated. In every 11 FEM simulation drop cases, FDR is 1 even the fuel assemblies are located inside of the cask. It is a quite larger value for all cases than the results with similar drop condition from the reports which covers the PRA on cask storage system. Because different from previous reports, subsequent impact was considered. Like in figure 8, accelerations which are used to calculate the FDR has extremely higher values in subsequent impact than the first impact for all SNF assemblies.

  5. Performance assessment analyses unique to Department of Energy spent nuclear fuel

    Loo, H.H.; Duguid, J.J.

    2000-01-01

    This paper describes the iterative process of grouping and performance assessment that has led to the current grouping of the U.S. Department of Energy (DOE) spent nuclear fuel (SNF). The unique sensitivity analyses that form the basis for incorporating DOE fuel into the total system performance assessment (TSPA) base case model are described. In addition, the chemistry that results from dissolution of DOE fuel and high level waste (HLW) glass in a failed co-disposal package, and the effects of disposal of selected DOE SNF in high integrity cans are presented

  6. Review of oxidation rates of DOE spent nuclear fuel : Part 1 : nuclear fuel

    Hilton, B.A.

    2000-01-01

    The long-term performance of Department of Energy (DOE) spent nuclear fuel (SNF) in a mined geologic disposal system depends highly on fuel oxidation and subsequent radionuclide release. The oxidation rates of nuclear fuels are reviewed in this two-volume report to provide a baseline for comparison with release rate data and technical rationale for predicting general corrosion behavior of DOE SNF. The oxidation rates of nuclear fuels in the DOE SNF inventory were organized according to metallic, Part 1, and non-metallic, Part 2, spent nuclear fuels. This Part 1 of the report reviews the oxidation behavior of three fuel types prototypic of metallic fuel in the DOE SNF inventory: uranium metal, uranium alloys and aluminum-based dispersion fuels. The oxidation rates of these fuels were evaluated in oxygen, water vapor, and water. The water data were limited to pure water corrosion as this represents baseline corrosion kinetics. Since the oxidation processes and kinetics discussed in this report are limited to pure water, they are not directly applicable to corrosion rates of SNF in water chemistry that is significantly different (such as may occur in the repository). Linear kinetics adequately described the oxidation rates of metallic fuels in long-term corrosion. Temperature dependent oxidation rates were determined by linear regression analysis of the literature data. As expected the reaction rates of metallic fuels dramatically increase with temperature. The uranium metal and metal alloys have stronger temperature dependence than the aluminum dispersion fuels. The uranium metal/water reaction exhibited the highest oxidation rate of the metallic fuel types and environments that were reviewed. Consequently, the corrosion properties of all DOE SNF may be conservatively modeled as uranium metal, which is representative of spent N-Reactor fuel. The reaction rate in anoxic, saturated water vapor was essentially the same as the water reaction rate. The long-term intrinsic

  7. Process to produce homogenized reactor fuels

    Hart, P.E.; Daniel, J.L.; Brite, D.W.

    1980-01-01

    The fuels consist of a mixture of PuO 2 and UO 2 . In order to increase the homogeneity of mechanically mixed fuels the pellets are sintered in a hydrogen atmosphere with a sufficiently low oxygen potential. This results in a reduction of Pu +4 to Pu +3 . By the reduction process water vapor is obtained increasing the pressure within the PuO 2 particles and causing PuO 2 to be pressed into the uranium oxide structure. (DG) [de

  8. Evaluation of Radionuclide Release from Aluminum-Based SNF in Basin Storage

    Sindelar, R.L.; Burke, S.D.; Howell, J.P.

    1998-09-01

    This report provides an evaluation of the release rates of radionuclides from breached A1-SNF assemblies and evaluates the effect of direct storage of breached fuel at a conservative upper bound reference condition on the SRS basin water activity levels

  9. Snf2 family gene distribution in higher plant genomes reveals DRD1 expansion and diversification in the tomato genome.

    Bargsten, Joachim W; Folta, Adam; Mlynárová, Ludmila; Nap, Jan-Peter

    2013-01-01

    As part of large protein complexes, Snf2 family ATPases are responsible for energy supply during chromatin remodeling, but the precise mechanism of action of many of these proteins is largely unknown. They influence many processes in plants, such as the response to environmental stress. This analysis is the first comprehensive study of Snf2 family ATPases in plants. We here present a comparative analysis of 1159 candidate plant Snf2 genes in 33 complete and annotated plant genomes, including two green algae. The number of Snf2 ATPases shows considerable variation across plant genomes (17-63 genes). The DRD1, Rad5/16 and Snf2 subfamily members occur most often. Detailed analysis of the plant-specific DRD1 subfamily in related plant genomes shows the occurrence of a complex series of evolutionary events. Notably tomato carries unexpected gene expansions of DRD1 gene members. Most of these genes are expressed in tomato, although at low levels and with distinct tissue or organ specificity. In contrast, the Snf2 subfamily genes tend to be expressed constitutively in tomato. The results underpin and extend the Snf2 subfamily classification, which could help to determine the various functional roles of Snf2 ATPases and to target environmental stress tolerance and yield in future breeding.

  10. Snf2 family gene distribution in higher plant genomes reveals DRD1 expansion and diversification in the tomato genome.

    Joachim W Bargsten

    Full Text Available As part of large protein complexes, Snf2 family ATPases are responsible for energy supply during chromatin remodeling, but the precise mechanism of action of many of these proteins is largely unknown. They influence many processes in plants, such as the response to environmental stress. This analysis is the first comprehensive study of Snf2 family ATPases in plants. We here present a comparative analysis of 1159 candidate plant Snf2 genes in 33 complete and annotated plant genomes, including two green algae. The number of Snf2 ATPases shows considerable variation across plant genomes (17-63 genes. The DRD1, Rad5/16 and Snf2 subfamily members occur most often. Detailed analysis of the plant-specific DRD1 subfamily in related plant genomes shows the occurrence of a complex series of evolutionary events. Notably tomato carries unexpected gene expansions of DRD1 gene members. Most of these genes are expressed in tomato, although at low levels and with distinct tissue or organ specificity. In contrast, the Snf2 subfamily genes tend to be expressed constitutively in tomato. The results underpin and extend the Snf2 subfamily classification, which could help to determine the various functional roles of Snf2 ATPases and to target environmental stress tolerance and yield in future breeding.

  11. Pyrochemical processing of DOE spent nuclear fuel

    Laidler, J.J.

    1995-01-01

    A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or open-quotes pyroprocessing,close quotes provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the United States Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (>99.9%) separation of transuranics. The resultant waste forms from the pyroprocess, are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory

  12. Regulation at nuclear fuel cycle

    2002-01-01

    This bulletin contains information about activities of the Nuclear Regulatory Authority of the Slovak Republic (UJD). In this leaflet the role of the UJD in regulation at nuclear fuel cycle is presented. The Nuclear Fuel Cycle (NFC) is a complex of activities linked with production of nuclear fuel for nuclear reactors as a source of energy used for production of electricity and heat, and of activities linked with spent nuclear fuel handling. Activities linked with nuclear fuel (NF) production, known as the Front-End of Nuclear Fuel Cycle, include (production of nuclear fuel from uranium as the most frequently used element). After discharging spent nuclear fuel (SNF) from nuclear reactor the activities follow linked with its storage, reprocessing and disposal known as the Back-End of Nuclear Fuel Cycle. Individual activity, which penetrates throughout the NFC, is transport of nuclear materials various forms during NF production and transport of NF and SNF. Nuclear reactors are installed in the Slovak Republic only in commercial nuclear power plants and the NFC is of the open type is imported from abroad and SNF is long-term supposed without reprocessing. The main mission of the area of NFC is supervision over: - assurance of nuclear safety throughout all NFC activities; - observance of provisions of the Treaty on Non-Proliferation of Nuclear Weapons during nuclear material handling; with an aim to prevent leakage of radioactive substances into environment (including deliberated danage of NFC sensitive facilities and misuse of nuclear materials to production of nuclear weapons. The UJD carries out this mission through: - assessment of safety documentation submitted by operators of nuclear installations at which nuclear material, NF and SNF is handled; - inspections concentrated on assurance of compliance of real conditions in NFC, i.e. storage and transport of NF and SNF; storage, transport and disposal of wastes from processing of SNF; with assumptions of the safety

  13. Development of advanced spent fuel management process. System analysis of advanced spent fuel management process

    Ro, S.G.; Kang, D.S.; Seo, C.S.; Lee, H.H.; Shin, Y.J.; Park, S.W.

    1999-03-01

    The system analysis of an advanced spent fuel management process to establish a non-proliferation model for the long-term spent fuel management is performed by comparing the several dry processes, such as a salt transport process, a lithium process, the IFR process developed in America, and DDP developed in Russia. In our system analysis, the non-proliferation concept is focused on the separation factor between uranium and plutonium and decontamination factors of products in each process, and the non-proliferation model for the long-term spent fuel management has finally been introduced. (Author). 29 refs., 17 tabs., 12 figs

  14. Method for processing spent nuclear reactor fuel

    Levenson, M.; Zebroski, E.L.

    1981-01-01

    A method and apparatus are claimed for processing spent nuclear reactor fuel wherein plutonium is continuously contaminated with radioactive fission products and diluted with uranium. Plutonium of sufficient purity to fabricate nuclear weapons cannot be produced by the process or in the disclosed reprocessing plant. Diversion of plutonium is prevented by radiation hazards and ease of detection

  15. Regulatory Experiences from Effective Step-wise Implementation of the SNF Disposal in Finland

    Hämäläinen, K.

    2016-01-01

    Finland is one of the foremost countries in the world in developing a disposal solution for spent nuclear fuel (SNF). The Construction License Application (CLA) for the Olkiluoto SNF encapsulation and disposal facility was submitted by Posiva, the implementer, to the authorities at the end of 2012 and the Government is expected to decide about the license during autumn 2015. In 1983 the Government made a strategy decision on the objectives and target time schedule for the research, development and technical planning of nuclear waste management. Decision included the milestones for site selection, submittal of construction license and start of disposal operations.

  16. Evaluation of Nondestructive Assay/Nondestructive Examination Capabilities for Department of Energy Spent Nuclear Fuel

    Luptak, A.J.; Bulmahn, K.D.

    1998-01-01

    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

  17. Brazing process in nuclear fuel element fabrication

    Katam, K.; Sudarsono

    1982-01-01

    The purpose of the brazing process is to join the spacers and pads of fuel pins, so that the process is meant as a soldering technique and not only as a hardening or reinforcing process such as in common brazing purposes. There are some preliminary processes before executing the brazing process such as: materials preparation, sand blasting, brazing metal coating tack welding the spacers and pads on the fuel cladding. The metal brazing used is beryllium in strip form which will be evaporated in vacuum condition to coat the spacers and pads. The beryllium vapor and dust is very hazardous to the workers, so all the line process of brazing needs specials safety protection and equipment to protect the workers and the processing area. Coating process temperature is 2470 deg C with a vacuum pressure of 10 -5 mmHg. Brazing process temperature process is 1060 deg C with a vacuum pressure of 10 -6 mmHg. The brazing process with beryllium coating probably will give metallurgical structural change in the fuel cladding metal at the locations of spacers and pads. The quality of brazing is highly influenced by and is depending on the chemical composition of the metal and the brazing metal, materials preparations, temperature, vacuum pressure, time of coating and brazing process. The quality control of brazing could be performed with methods of visuality geometry, radiography and metallography. (author)

  18. An innovative way of thinking nuclear waste management - Neutron physics of a reactor directly operating on SNF.

    Merk, Bruno; Litskevich, Dzianis; Bankhead, Mark; Taylor, Richard J

    2017-01-01

    A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T) promises a solution for improved waste management. Current strategies rely on systems designed in the 60's for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF) without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient.

  19. Cost analysis of spent nuclear fuel management

    Robertson, D.L.M.; Ford, L.M.

    1993-01-01

    The Department of Energy Civilian Radioactive Waste Management System (CRWMS) is chartered to develop a waste management system for the safe disposal of spent nuclear fuel (SNF) from the 131 nuclear power reactors in the United States and a certain amount of high level waste (HLW) from reprocessing operations. The current schedule is to begin accepting SNF in 1998 for storage at a Monitored Retrievable Storage (MRS) facility. Subsequently, beginning in 2010, the system is scheduled to begin accepting SNF at a permanent geologic repository in 2010 and HLW in 2015. At this time, a MRS site has not been selected. Yucca Mountain, Nevada is currently being evaluated as the candidate site for the repository for permanent geologic disposal of SNF. All SNF, with the possible exception of the SNF from the western reactors, is currently planned to be shipped to or through the MRS site en route to the repository. The repository will operate in an acceptance and performance confirmation phase for a 50 year period beginning in 2010 with an additional nine year closure and five year decontamination and decommissioning period. The MRS has a statutory maximum capacity of 15,000 Metric Tons Uranium (MTU), with a further restriction that it may not store more than 10,000 MTU until the repository begins accepting waste. The repository is currently scheduled to store 63,000 MTU of SNF and an additional 7,000 MTU equivalent of HLW for a total capacity of 70,000 MTU. The amended act specified the MRS storage limits and identified Yucca Mountain as the only site to be characterized. Also, an Office of the Nuclear Waste Negotiator was established to secure a voluntary host site for the MRS. The MRS, the repository, and all waste containers/casks will go through a Nuclear Regulatory Commission licensing process much like the licensing process for a nuclear power plant. Environmental assessments and impact statements will be prepared for both the MRS and repository

  20. Development of advanced spent fuel management process

    Park, Seong Won; Shin, Y. J.; Cho, S. H.

    2004-03-01

    The research on spent fuel management focuses on the maximization of the disposal efficiency by a volume reduction, the improvement of the environmental friendliness by the partitioning and transmutation of the long lived nuclides, and the recycling of the spent fuel for an efficient utilization of the uranium source. In the second phase which started in 2001, the performance test of the advanced spent fuel management process consisting of voloxidation, reduction of spent fuel and the lithium recovery process has been completed successfully on a laboratory scale. The world-premier spent fuel reduction hot test of a 5 kgHM/batch has been performed successfully by joint research with Russia and the valuable data on the actinides and FPs material balance and the characteristics of the metal product were obtained with experience to help design an engineering scale reduction system. The electrolytic reduction technology which integrates uranium oxide reduction in a molten LiCl-Li 2 O system and Li 2 O electrolysis is developed and a unique reaction system is also devised. Design data such as the treatment capacity, current density and mass transfer behavior obtained from the performance test of a 5 kgU/batch electrolytic reduction system pave the way for the third phase of the hot cell demonstration of the advanced spent fuel management technology

  1. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    Barkauskas, V., E-mail: vytenis.barkauskas@ftmc.lt; Plukiene, R., E-mail: rita.plukiene@ftmc.lt; Plukis, A., E-mail: arturas.plukis@ftmc.lt

    2016-10-15

    Highlights: • RBMK-1500 fuel burn-up impact on k{sub eff} in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k{sub eff} in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k{sub eff}) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality

  2. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    Barkauskas, V.; Plukiene, R.; Plukis, A.

    2016-01-01

    Highlights: • RBMK-1500 fuel burn-up impact on k_e_f_f in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k_e_f_f in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k_e_f_f) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality safety.

  3. COMMERCIAL SNF ACCIDENT RELEASE FRACTIONS

    S.O. Bader

    1999-10-18

    The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. The radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses. This subset of the total release fraction is referred to as the respirable release fraction. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be

  4. COMMERCIAL SNF ACCIDENT RELEASE FRACTIONS

    S.O. Bader

    1999-01-01

    The purpose of this design analysis is to specify and document the total and respirable fractions for radioactive materials that are released from an accident event at the Monitored Geologic Repository (MGR) involving commercial spent nuclear fuel (CSNF) in a dry environment. The total and respirable release fractions will be used to support the preclosure licensing basis for the MGR. The total release fraction is defined as the fraction of total CSNF assembly inventory, typically expressed as an activity inventory (e.g., curies), of a given radionuclide that is released to the environment from a waste form. The radionuclides are released from the inside of breached fuel rods (or pins) and from the detachment of radioactive material (crud) from the outside surfaces of fuel rods and other components of fuel assemblies. The total release fraction accounts for several mechanisms that tend to retain, retard, or diminish the amount of radionuclides that are available for transport to dose receptors or otherwise can be shown to reduce exposure of receptors to radiological releases. The total release fraction includes a fraction of airborne material that is respirable and could result in inhalation doses. This subset of the total release fraction is referred to as the respirable release fraction. Potential accidents may involve waste forms that are characterized as either bare (unconfined) fuel assemblies or confined fuel assemblies. The confined CSNF assemblies at the MGR are contained in shipping casks, canisters, or disposal containers (waste packages). In contrast to the bare fuel assemblies, the container that confines the fuel assemblies has the potential of providing an additional barrier for diminishing the total release fraction should the fuel rod cladding breach during an accident. However, this analysis will not take credit for this additional bamer and will establish only the total release fractions for bare unconfined CSNF assemblies, which may however be

  5. Development of advanced spent fuel management process

    Shin, Young Joon; Cho, S. H.; You, G. S.

    2001-04-01

    Currently, the economic advantage of any known approach to the back end fuel cycle of a nuclear power reactor has not been well established. Thus the long term storage of the spent fuel in a safe manner is one of the important issues to be resolved in countries where the nuclear power has a relatively heavy weight in power production of that country. At KAERI, as a solution to this particular issue midterm storage of the spent fuel, an alternative approach has been developed. This approach includes the decladding and pulverization process of the spent PWR fuel rod, the reducing process from the uranium oxide to a metallic uranium powder using Li metal in a LiCl salt, the continuous casting process of the reduced metal, and the recovery process of Li from mixed salts by the electrolysis. We conducted the laboratory scale tests of each processes for the technical feasibility and determination for the operational conditions for this approach. Also, we performed the theoretical safety analysis and conducted integral tests for the equipment integration through the Mock-up facility with non-radioactive samples. There were no major issues in the approach, however, material incompatibility of the alkaline metal and oxide in a salt at a high temperature and the reactor that contains the salt became a show stopper of the process. Also the difficulty of the clear separation of the salt with metals reduced from the oxide became a major issue

  6. Development of advanced spent fuel management process

    Ro, Seung Gy; Shin, Y. J.; Do, J. B.; You, G. S.; Seo, J. S.; Lee, H. G.

    1998-03-01

    This study is to develop an advanced spent fuel management process for countries which have not yet decided a back-end nuclear fuel cycle policy. The aims of this process development based on the pyroreduction technology of PWR spent fuels with molten lithium, are to reduce the storage volume by a quarter and to reduce the storage cooling load in half by the preferential removal of highly radioactive decay-heat elements such as Cs-137 and Sr-90 only. From the experimental results which confirm the feasibility of metallization technology, it is concluded that there are no problems in aspects of reaction kinetics and equilibrium. However, the operating performance test of each equipment on an engineering scale still remain and will be conducted in 1999. (author). 21 refs., 45 tabs., 119 figs

  7. Review on Fuel Loading Process and Performance for Advanced Fuel Handling Equipment

    Chang, Sang-Gyoon; Lee, Dae-Hee; Kim, Young-Baik; Lee, Deuck-Soo

    2007-01-01

    The fuel loading process and the performance of the advanced fuel handling equipment for OPR 1000 (Optimized Power Plant) are analyzed and evaluated. The fuel handling equipment, which acts critical processes in the refueling outage, has been improved to reduce fuel handling time. The analysis of the fuel loading process can be a useful tool to improve the performance of the fuel handling equipment effectively. Some recommendations for further improvement are provided based on this study

  8. Alternative Fuel for Portland Cement Processing

    Schindler, Anton K; Duke, Steve R; Burch, Thomas E; Davis, Edward W; Zee, Ralph H; Bransby, David I; Hopkins, Carla; Thompson, Rutherford L; Duan, Jingran; ; Venkatasubramanian, Vignesh; Stephen, Giles

    2012-06-30

    The production of cement involves a combination of numerous raw materials, strictly monitored system processes, and temperatures on the order of 1500 °C. Immense quantities of fuel are required for the production of cement. Traditionally, energy from fossil fuels was solely relied upon for the production of cement. The overarching project objective is to evaluate the use of alternative fuels to lessen the dependence on non-renewable resources to produce portland cement. The key objective of using alternative fuels is to continue to produce high-quality cement while decreasing the use of non-renewable fuels and minimizing the impact on the environment. Burn characteristics and thermodynamic parameters were evaluated with a laboratory burn simulator under conditions that mimic those in the preheater where the fuels are brought into a cement plant. A drop-tube furnace and visualization method were developed that show potential for evaluating time- and space-resolved temperature distributions for fuel solid particles and liquid droplets undergoing combustion in various combustion atmospheres. Downdraft gasification has been explored as a means to extract chemical energy from poultry litter while limiting the throughput of potentially deleterious components with regards to use in firing a cement kiln. Results have shown that the clinkering is temperature independent, at least within the controllable temperature range. Limestone also had only a slight effect on the fusion when used to coat the pellets. However, limestone addition did display some promise in regards to chlorine capture, as ash analyses showed chlorine concentrations of more than four times greater in the limestone infused ash as compared to raw poultry litter. A reliable and convenient sampling procedure was developed to estimate the combustion quality of broiler litter that is the best compromise between convenience and reliability by means of statistical analysis. Multi-day trial burns were conducted

  9. Process of producing a fuel, etc

    1924-12-01

    This invention has for its object a process of producing fuels by separating a light oil from primary tar, characterized by a succession of operations comprising preliminary removal of phenols from the oils, removing sulfur completely by the application of suitable catalysts and an agent to fix the free sulfur as hydrogen sulfide; finally, washing to remove ethylenes, pyridines, and impurities from the treatment.

  10. Spent Nuclear Fuel Project operational staffing plan

    Debban, B.L.

    1996-03-01

    Using the Spent Nuclear Fuel (SNF) Project's current process flow concepts and knowledge from cognizant engineering and operational personnel, an initial assessment of the SNF Project radiological exposure and resource requirements was completed. A small project team completed a step by step analysis of fuel movement in the K Basins to the new interim storage location, the Canister Storage Building (CSB). This analysis looked at fuel retrieval, conditioning of the fuel, and transportation of the fuel. This plan describes the staffing structure for fuel processing, fuel movement, and the maintenance and operation (M ampersand O) staffing requirements of the facilities. This initial draft does not identify the support function resources required for M ampersand O, i.e., administrative and engineering (technical support). These will be included in future revisions to the plan. This plan looks at the resource requirements for the SNF subprojects, specifically, the operations of the facilities, balances resources where applicable, rotates crews where applicable, and attempts to use individuals in multi-task assignments. This plan does not apply to the construction phase of planned projects that affect staffing levels of K Basins

  11. K Basin Fuel Characterization Program Technical Baseline Summary

    SUYAMA, R.M.

    1999-01-01

    This document provides a summary of the systematic process used by the SNF Project to characterize K-Basin spent fuel, and to develop and apply the appropriate conservative safety margins to the resulting parameters for technical designs and safety analyses

  12. Process and device to produce fuel briquettes

    Caroe, C J

    1980-10-23

    A two-stage process for the production of briquettes consisting essentially of cellulose (sawdust, peanut shells) is proposed. The fuel material (in case with additives) is molded by high pressure to pellets of the size of a few centimeters. The pellets are mixed with flammable binding agents like paraffin, wax, polyethylene etc. and molded at a lower pressure or extruded in a second step. A suited molding device is described. The wax content could be lowered with respect to known processes.

  13. Will the world SNF be reprocessed in Russia?

    Gagarinski, A.

    2000-01-01

    Russia's possibilities in nuclear fuel reprocessing are well known. RT-1 plant with 400 tons/year in the Chelyabinsk region can provide reprocessing of fuel from Russian and Central European WWER-440 reactors, as well as from transport and research reactors. Former military complex Krasnoyarsk-26 with unique underground installations situated in rock galleries, already has an aqueous facility for storage of 6000 tons of spent nuclear fuel (SNF), half-built plant RT-2 for nuclear fuel reprocessing with 1500 tons/year capacity, as well as the projects of dry storage facility for 30000 tons of SNF and of MOX fuel production plant. Russian nuclear specialists understand well, that the economic efficiency of nuclear fuel reprocessing industry is shown only in case of large-scale production, which would require consolidation of the countries, which develop nuclear energy. They also understand, that Russia has all the possibilities to become one of the centers of such a consolidation and to use these possibilities for the benefit of the country. The idea of foreign nuclear fuel reprocessing (for a long time realized for East and Central European countries, which operate Soviet-design reactors) has existed in the specialists' minds, and sometimes has appeared in the mass media. On the other hand, rehabilitation of territories of nuclear fuel cycle enterprises in Russia continues, including the Karachai lake, which contains 120 million Curie of radioactivity. Unfortunately, Russia simply has no money for complete solution of the problems of radiation military legacy. During discussion of the budget for 2000, the Russian Minatom has made a daring step. A real program, how to find money needed for solving the 'radiation legacy' problem, was proposed. With this purpose, it was proposed to permit storage and further reprocessing of other countries' SNF on Russian territory. It is well known, that another countries' SNF is accepted for reprocessing by UK and France, and Russia

  14. Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

    Bibilashvili, Yu K; Nekrasova, G A; Sukhanov, G I

    1989-01-01

    In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified.

  15. Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

    Bibilashvili, Yu.K.; Nekrasova, G.A.; Sukhanov, G.I.

    1989-01-01

    In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified

  16. Safeguardability of advanced spent fuel conditioning process

    Li, T. K. (Tien K.); Lee, S. Y. (Sang Yoon); Burr, Tom; Russo, P. A. (Phyllis A.); Menlove, Howard O.; Kim, H. D.; Ko, W. I. (Won Il); Park, S. W.; Park, H. S.

    2004-01-01

    The Advanced Spent Fuel Conditioning Process (ACP) is an electro-metallurgical treatment technique to convert oxide-type spent nuclear fuel into a metallic form. The Korea Atomic Energy Research Institute (KAERI) has been developing this technology since 1977 for the purpose of spent fuel management and is planning to perform a lab-scale demonstration in 2006. By using of this technology, a significant reduction of the volume and heat load of spent fuel is expected, which would lighten the burden of final disposal in terms of disposal size, safety and economics. In the framework of collaboration agreement to develop the safeguards system for the ACP, a joint study on the safeguardability of the ACP technology has been performed by the Los Alamos National Laboratory (LANL) and the KAERI since 2002. In this study, the safeguardability of the ACP technology was examined for the pilot-scale facility. The process and material flows were conceptually designed, and the uncertainties in material accounting were estimated with international target values.

  17. Spent Nuclear Fuel Alternative Technology Risk Assessment

    Perella, V.F.

    1999-11-29

    A Research Reactor Spent Nuclear Fuel Task Team (RRTT) was chartered by the Department of Energy (DOE) Office of Spent Fuel Management with the responsibility to recommend a course of action leading to a final technology selection for the interim management and ultimate disposition of the foreign and domestic aluminum-based research reactor spent nuclear fuel (SNF) under DOE''s jurisdiction. The RRTT evaluated eleven potential SNF management technologies and recommended that two technologies, direct co-disposal and an isotopic dilution alternative, either press and dilute or melt and dilute, be developed in parallel. Based upon that recommendation, the Westinghouse Savannah River Company (WSRC) organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and provide a WSRC recommendation to DOE for a preferred SNF alternative management technology. A technology risk assessment was conducted as a first step in this recommendation process to determine if either, or both, of the technologies posed significant risks that would make them unsuitable for further development. This report provides the results of that technology risk assessment.

  18. Spent Nuclear Fuel Alternative Technology Risk Assessment

    Perella, V.F.

    1999-01-01

    A Research Reactor Spent Nuclear Fuel Task Team (RRTT) was chartered by the Department of Energy (DOE) Office of Spent Fuel Management with the responsibility to recommend a course of action leading to a final technology selection for the interim management and ultimate disposition of the foreign and domestic aluminum-based research reactor spent nuclear fuel (SNF) under DOE''s jurisdiction. The RRTT evaluated eleven potential SNF management technologies and recommended that two technologies, direct co-disposal and an isotopic dilution alternative, either press and dilute or melt and dilute, be developed in parallel. Based upon that recommendation, the Westinghouse Savannah River Company (WSRC) organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and provide a WSRC recommendation to DOE for a preferred SNF alternative management technology. A technology risk assessment was conducted as a first step in this recommendation process to determine if either, or both, of the technologies posed significant risks that would make them unsuitable for further development. This report provides the results of that technology risk assessment

  19. A multi-attribute utility decision analysis for treatment alternatives for the DOE/SR aluminum-based spent nuclear fuel

    Davis, Freddie J.; Weiner, Ruth Fleischman; Wheeler, Timothy A.; Sorenson, Ken B.; Kuzio, Kenneth A.

    2000-01-01

    A multi-attribute utility analysis is applied to a decision process to select a treatment method for the management of aluminum-based spent nuclear fuel (Al-SNF) owned by the US 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-SNF to a geologic repository for permanent disposal. DOE initially considered ten treatment alternatives for the management of Al-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 formal decision process used to evaluate these two remaining alternatives

  20. Process development and fabrication for sphere-pac fuel rods

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted

  1. 78 FR 20625 - Spent Nuclear Fuel Management at the Savannah River Site

    2013-04-05

    ... Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact... generated at the Oak Ridge National Laboratory and approximately 1,000 bundles of aluminum-clad SNF... processing is a chemical separations process that involves dissolving spent fuel in nitric acid and...

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

    Abrefah, J.; Gray, W.J.; Ketner, G.L.; Marschman, S.C.; Pyecha, T.D.; Thornton, T.A.

    1996-03-01

    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

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

    Abrefah, J.; Gray, W.J.; Ketner, G.L.; Marschman, S.C.; Pyecha, T.D.; Thornton, T.A.

    1996-03-01

    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.

  4. Modeling closed nuclear fuel cycles processes

    Shmidt, O.V. [A.A. Bochvar All-Russian Scientific Research Institute for Inorganic Materials, Rogova, 5a street, Moscow, 123098 (Russian Federation); Makeeva, I.R. [Zababakhin All-Russian Scientific Research Institute of Technical Physics, Vasiliev street 13, Snezhinsk, Chelyabinsk region, 456770 (Russian Federation); Liventsov, S.N. [Tomsk Polytechnic University, Tomsk, Lenin Avenue, 30, 634050 (Russian Federation)

    2016-07-01

    Computer models of processes are necessary for determination of optimal operating conditions for closed nuclear fuel cycle (NFC) processes. Computer models can be quickly changed in accordance with new and fresh data from experimental research. 3 kinds of process simulation are necessary. First, the VIZART software package is a balance model development used for calculating the material flow in technological processes. VIZART involves taking into account of equipment capacity, transport lines and storage volumes. Secondly, it is necessary to simulate the physico-chemical processes that are involved in the closure of NFC. The third kind of simulation is the development of software that allows the optimization, diagnostics and control of the processes which implies real-time simulation of product flows on the whole plant or on separate lines of the plant. (A.C.)

  5. Industrial Maturity of FR Fuel Cycle Processes and Technologies

    Bruezière, Jérôme

    2013-01-01

    FR fuel cycle processes and technologies have already been proven industrially for Oxide Fuel, and to a lesser extent for metal fuel. In addition, both used oxide fuel reprocessing and fresh oxide fuel manufacturing benefit from similar industrial experience currently deployed for LWR. Alternative fuel type will have to generate very significant benefit in reactor ( safety, cost, … ) to justify corresponding development and industrialization costs

  6. Modification in fuel processing of Mitsubishi Nuclear Fuel's Tokai Works

    1976-01-01

    Results of the study by the Committee for Examination of Fuel Safety, reported to the AEC of Japan, are presented, concerning safety of the modifications of Tokai Works, Mitsubishi Nuclear Fuel Co., Ltd. Safety has been confirmed thereof. The modifications covered are the following: storage facility of nuclear fuel in increase, analytical facility in transfer, fuel assemblage equipment in addition, incineration facility of combustible solid wastes in installation, experimental facility of uranium recovery in installation, and warehouse in installation. (Mori, K.)

  7. Identification of multiple distinct Snf2 subfamilies with conserved structural motifs.

    Flaus, Andrew; Martin, David M A; Barton, Geoffrey J; Owen-Hughes, Tom

    2006-01-01

    The Snf2 family of helicase-related proteins includes the catalytic subunits of ATP-dependent chromatin remodelling complexes found in all eukaryotes. These act to regulate the structure and dynamic properties of chromatin and so influence a broad range of nuclear processes. We have exploited progress in genome sequencing to assemble a comprehensive catalogue of over 1300 Snf2 family members. Multiple sequence alignment of the helicase-related regions enables 24 distinct subfamilies to be identified, a considerable expansion over earlier surveys. Where information is known, there is a good correlation between biological or biochemical function and these assignments, suggesting Snf2 family motor domains are tuned for specific tasks. Scanning of complete genomes reveals all eukaryotes contain members of multiple subfamilies, whereas they are less common and not ubiquitous in eubacteria or archaea. The large sample of Snf2 proteins enables additional distinguishing conserved sequence blocks within the helicase-like motor to be identified. The establishment of a phylogeny for Snf2 proteins provides an opportunity to make informed assignments of function, and the identification of conserved motifs provides a framework for understanding the mechanisms by which these proteins function.

  8. Mechanisms of regulation of SNF1/AMPK/SnRK1 protein kinases

    Crozet, Pierre; Margalha, Leonor; Confraria, Ana; Rodrigues, Américo; Martinho, Cláudia; Adamo, Mattia; Elias, Carlos A.; Baena-González, Elena

    2014-01-01

    The SNF1 (sucrose non-fermenting 1)-related protein kinases 1 (SnRKs1) are the plant orthologs of the budding yeast SNF1 and mammalian AMPK (AMP-activated protein kinase). These evolutionarily conserved kinases are metabolic sensors that undergo activation in response to declining energy levels. Upon activation, SNF1/AMPK/SnRK1 kinases trigger a vast transcriptional and metabolic reprograming that restores energy homeostasis and promotes tolerance to adverse conditions, partly through an induction of catabolic processes and a general repression of anabolism. These kinases typically function as a heterotrimeric complex composed of two regulatory subunits, β and γ, and an α-catalytic subunit, which requires phosphorylation of a conserved activation loop residue for activity. Additionally, SNF1/AMPK/SnRK1 kinases are controlled by multiple mechanisms that have an impact on kinase activity, stability, and/or subcellular localization. Here we will review current knowledge on the regulation of SNF1/AMPK/SnRK1 by upstream components, post-translational modifications, various metabolites, hormones, and others, in an attempt to highlight both the commonalities of these essential eukaryotic kinases and the divergences that have evolved to cope with the particularities of each one of these systems. PMID:24904600

  9. DAF-16 employs the chromatin remodeller SWI/SNF to promote stress resistance and longevity.

    Riedel, Christian G; Dowen, Robert H; Lourenco, Guinevere F; Kirienko, Natalia V; Heimbucher, Thomas; West, Jason A; Bowman, Sarah K; Kingston, Robert E; Dillin, Andrew; Asara, John M; Ruvkun, Gary

    2013-05-01

    Organisms are constantly challenged by stresses and privations and require adaptive responses for their survival. The forkhead box O (FOXO) transcription factor DAF-16 (hereafter referred to as DAF-16/FOXO) is a central nexus in these responses, but despite its importance little is known about how it regulates its target genes. Proteomic identification of DAF-16/FOXO-binding partners in Caenorhabditis elegans and their subsequent functional evaluation by RNA interference revealed several candidate DAF-16/FOXO cofactors, most notably the chromatin remodeller SWI/SNF. DAF-16/FOXO and SWI/SNF form a complex and globally co-localize at DAF-16/FOXO target promoters. We show that specifically for gene activation, DAF-16/FOXO depends on SWI/SNF, facilitating SWI/SNF recruitment to target promoters, to activate transcription by presumed remodelling of local chromatin. For the animal, this translates into an essential role for SWI/SNF in DAF-16/FOXO-mediated processes, in particular dauer formation, stress resistance and the promotion of longevity. Thus, we give insight into the mechanisms of DAF-16/FOXO-mediated transcriptional regulation and establish a critical link between ATP-dependent chromatin remodelling and lifespan regulation.

  10. DAF-16/FOXO employs the chromatin remodeller SWI/SNF to promote stress resistance and longevity

    Riedel, Christian G.; Dowen, Robert H.; Lourenco, Guinevere F.; Kirienko, Natalia V.; Heimbucher, Thomas; West, Jason A.; Bowman, Sarah K.; Kingston, Robert E.; Dillin, Andrew; Asara, John M.; Ruvkun, Gary

    2013-01-01

    Organisms are constantly challenged by stresses and privations and require adaptive responses for their survival. The transcription factor DAF-16/FOXO is central nexus in these responses, but despite its importance little is known about how it regulates its target genes. Proteomic identification of DAF-16/FOXO binding partners in Caenorhabditis elegans and their subsequent functional evaluation by RNA interference (RNAi) revealed several candidate DAF-16/FOXO cofactors, most notably the chromatin remodeller SWI/SNF. DAF-16/FOXO and SWI/SNF form a complex and globally colocalize at DAF-16/FOXO target promoters. We show that specifically for gene-activation, DAF-16/FOXO depends on SWI/SNF, facilitating SWI/SNF recruitment to target promoters, in order to activate transcription by presumed remodelling of local chromatin. For the animal, this translates into an essential role of SWI/SNF for DAF-16/FOXO-mediated processes, i.e. dauer formation, stress resistance, and the promotion of longevity. Thus we give insight into the mechanisms of DAF-16/FOXO-mediated transcriptional regulation and establish a critical link between ATP-dependent chromatin remodelling and lifespan regulation. PMID:23604319

  11. Functional identification of an Arabidopsis snf4 ortholog by screening for heterologous multicopy suppressors of snf4 deficiency in yeast

    Kleinow, T.; Bhalerao, R.; Breuer, F.

    2000-01-01

    Yeast Snf4 is a prototype of activating gamma-subunits of conserved Snf1/AMPK-related protein kinases (SnRKs) controlling glucose and stress signaling in eukaryotes. The catalytic subunits of Arabidopsis SnRKs, AKIN10 and AKIN11, interact with Snf4 and suppress the snf1 and snf4 mutations in yeast....... By expression of an Arabidopsis cDNA library in yeast, heterologous multicopy snf4 suppressors were isolated. In addition to AKIN10 and AKIN11, the deficiency of yeast snf4 mutant to grown on non-fermentable carbon source was suppressed by Arabidopsis Myb30, CAAT-binding factor Hap3b, casein kinase I, zinc......-finger factors AZF2 and ZAT10, as well as orthologs of hexose/UDP-hexose transporters, calmodulin, SMC1-cohesin and Snf4. Here we describe the characterization of AtSNF4, a functional Arabidopsis Snf4 ortholog, that interacts with yeast Snf1 and specifically binds to the C-terminal regulatory domain...

  12. SNF Interim Storage Canister Corrosion and Surface Environment Investigations

    Bryan, Charles R.; Enos, David G.

    2015-01-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In order for SCC to occur, three criteria must be met. A corrosive environment must be present on the canister surface, the metal must susceptible to SCC, and sufficient tensile stress to support SCC must be present through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.

  13. SNF Interim Storage Canister Corrosion and Surface Environment Investigations

    Bryan, Charles R. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Enos, David G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In order for SCC to occur, three criteria must be met. A corrosive environment must be present on the canister surface, the metal must susceptible to SCC, and sufficient tensile stress to support SCC must be present through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.

  14. Pyroelectrochemical process for reprocessing irradiated nuclear fuels

    Brambilla, G.; Sartorelli, A.

    1982-01-01

    A pyroelectrochemical process for reprocessing irradiated fast reactor mixed oxide or carbide fuels is described. The fuel is dissolved in a bath of molten alkali metal sulfates. The Pu(SO 4 ) 2 formed in the bath is thermally decomposed, leaving crystalline PuO 2 on the bottom of the reaction vessel. Electrodes are then introduced into the bath, and UO 2 is deposited on the cathode. Alternatively, both UO 2 and PuO 2 may be electrodeposited. The molten salts, after decontamination by precipitating the fission products dissolved in the bath by introducing basic agents such as oxides, carbonates, or hydroxides, may be recycled. Since it is not possible to remove cesium from the molten salt bath, periodic disposal and partial renewal with fresh salts is necessary. The melted salts that contain the fission products are conditioned for disposal by embedding them in a metallic matrix

  15. Plasma Technologies of Solid Fuels Processing

    Karpenko, E.I.; Messerle, V.E.; Ustimenko, A.

    2003-01-01

    Use of fuel processing plasma technologies improves ecological and economical indexes of low-grade coal utilization at thermal power plants. This paper presents experimental plasma plant 70 k W of power and 11 kg per hour of coal productivity. On the base of material and heat balances integral indexes of the process of plasma gasification of Podmoskovny brown coal 48% of ash content were found. Synthesis gas with concentration 85.2% was got. Hydrogen concentration in the synthesis gas was higher than carbon monoxide one. Ratio H 2 :CO in synthesis gas was 1.4-1.5. It was shown that steam consumption and temperature of the process increase causes H 2 concentration and coal gasification degree increase. Fulfilled experiments and comparison of their result with theoretical investigations allowed creating pilot experimental plant for plasma processing of low-grade coals. The power of the pilot plant is 1000 k W and coal productivity is 300 kg/h. (author)

  16. Chemical process safety at fuel cycle facilities

    Ayres, D.A.

    1997-08-01

    This NUREG provides broad guidance on chemical safety issues relevant to fuel cycle facilities. It describes an approach acceptable to the NRC staff, with examples that are not exhaustive, for addressing chemical process safety in the safe storage, handling, and processing of licensed nuclear material. It expounds to license holders and applicants a general philosophy of the role of chemical process safety with respect to NRC-licensed materials; sets forth the basic information needed to properly evaluate chemical process safety; and describes plausible methods of identifying and evaluating chemical hazards and assessing the adequacy of the chemical safety of the proposed equipment and facilities. Examples of equipment and methods commonly used to prevent and/or mitigate the consequences of chemical incidents are discussed in this document

  17. WWER-1000 nuclear fuel manufacturing process at PJSC MSZ

    Morylev, A.; Bagdatyeva, E.; Aksenov, P.

    2015-01-01

    In this report a brief description of WWER-1000 fuel manufacturing process steps at PJSC MSZ as: uranium dioxide powder fabrication; fuel pellet manufacture fuel rod manufacture working assembly and fuel assembly manufacture is given. The implemented innovations are also presented

  18. Criticality Potential of Waste Packages Containing DOE SNF Affected by Igneous Intrusion

    D.S. Kimball; C.E. Sanders

    2006-01-01

    The Department of Energy (DOE) is currently preparing an application to submit to the U.S. Nuclear Regulatory Commission for a construction authorization for a monitored geologic repository. The repository will contain spent nuclear fuel (SNF) and defense high-level waste (DHLW) in waste packages placed in underground tunnels, or drifts. The primary objective of this paper is to perform a criticality analysis for waste packages containing DOE SNF affected by a disruptive igneous intrusion event in the emplacement drifts. The waste packages feature one DOE SNF canister placed in the center and surrounded by five High-Level Waste (HLW) glass canisters. The effective neutron multiplication factor (k eff ) is determined for potential configurations of the waste package during and after an intrusive igneous event. Due to the complexity of the potential scenarios following an igneous intrusion, finding conservative and bounding configurations with respect to criticality requires some additional considerations. In particular, the geometry of a slumped and damaged waste package must be examined, drift conditions must be modeled over a range of parameters, and the chemical degradation of DOE SNF and waste package materials must be considered for the expected high temperatures. The secondary intent of this calculation is to present a method for selecting conservative and bounding configurations for a wide range of end conditions

  19. Complex plasmochemical processing of solid fuel

    Vladimir Messerle

    2012-12-01

    Full Text Available Technology of complex plasmaochemical processing of solid fuel by Ecibastuz bituminous and Turgay brown coals is presented. Thermodynamic and experimental study of the technology was fulfilled. Use of this technology allows producing of synthesis gas from organic mass of coal and valuable components (technical silicon, ferrosilicon, aluminum and silicon carbide and microelements of rare metals: uranium, molybdenum, vanadium etc. from mineral mass of coal. Produced a high-calorific synthesis gas can be used for methanol synthesis, as high-grade reducing gas instead of coke, as well as energy gas in thermal power plants.

  20. Direct conversion of surplus fissile materials, spent nuclear fuel, and other materials to high-level-waste glass

    Forsberg, C.W.; Elam, K.R.

    1995-01-01

    With the end of the cold war the United States, Russia, and other countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. The United States Academy of Sciences (NAS) has recommended that these surplus fissile materials (SFMs) be processed so they are no more accessible than plutonium in spent nuclear fuel (SNF). This spent fuel standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. The NAS recommended investigation of three sets of options for disposition of SFMs while meeting the spent fuel standard: (1) incorporate SFMs with highly radioactive materials and dispose of as waste, (2) partly burn the SFMs in reactors with conversion of the SFMs to SNF for disposal, and (3) dispose of the SFMs in deep boreholes. The US Government is investigating these options for SFM disposition. A new method for the disposition of SFMs is described herein: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptinium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal

  1. Interface agreement for the management of 308 Building Spent Nuclear Fuel. Revision 1

    Danko, A.D.

    1995-01-01

    The Hanford Site Spent Nuclear Fuel (SNF) Project was formed to manage the SNF at Hanford. Specifically, the mission of the SNF Project on the Hanford Site is to ''provide safe, economic, environmentally sound management of Hanford SNF in a manner which stages it for final disposition.'' The current mission of the Fuel Fabrication Facilities Transition Project (FFFTP) is to transition the 308 Building for turn over to the Environmental Restoration Contractor for decontamination and decommissioning

  2. Management Of Hanford KW Basin Knockout Pot Sludge As Spent Nuclear Fuel

    Raymond, R. E.; Evans, K. M.

    2012-01-01

    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)

  3. System Theoretic Frameworks for Mitigating Risk Complexity in the Nuclear Fuel Cycle

    Williams, Adam David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mohagheghi, Amir H. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cohn, Brian [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, Douglas M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Jones, Katherine A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); DeMenno, Mercy [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kalinina, Elena Arkadievna [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Thomas, Maikael A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Parks, Ethan Rutledge [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Parks, Mancel Jordan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Jeantete, Brian A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-01

    In response to the expansion of nuclear fuel cycle (NFC) activities -- and the associated suite of risks -- around the world, this project evaluated systems-based solutions for managing such risk complexity in multimodal and multi-jurisdictional international spent nuclear fuel (SNF) transportation. By better understanding systemic risks in SNF transportation, developing SNF transportation risk assessment frameworks, and evaluating these systems-based risk assessment frameworks, this research illustrated interdependency between safety, security, and safeguards risks is inherent in NFC activities and can go unidentified when each "S" is independently evaluated. Two novel system-theoretic analysis techniques -- dynamic probabilistic risk assessment (DPRA) and system-theoretic process analysis (STPA) -- provide integrated "3S" analysis to address these interdependencies and the research results suggest a need -- and provide a way -- to reprioritize United States engagement efforts to reduce global nuclear risks. Lastly, this research identifies areas where Sandia National Laboratories can spearhead technical advances to reduce global nuclear dangers.

  4. Dissolution process for advanced-PWR-type fuels

    Black, D.E.; Decker, L.A.; Pearson, L.G.

    1979-01-01

    The new Fluorinel Dissolution Process and Fuel Storage (FAST) Facility at ICPP will provide underwater storage of spent PWR fuel and a new head-end process for fuel dissolution. The dissolution will be two-stage, using HF and HNO 3 , with an intermittent H 2 SO 4 dissolution for removing stainless steel components. Equipment operation is described

  5. Demonstrating the Feasibility of Molten Aluminum for Destroying Polymeric Encapsulants in SNF-Bearing Metallographic Mounts. Final Technical Report

    Dan Stout; Scott Ploger

    2004-01-01

    DOE-owned spent nuclear fuel (SNF) rods have been cross sectioned and mounted for metallography throughout the history of nuclear reactors. Many hundreds of these ''met mounts'' have accumulated in storage across the DOE complex. However, because of potential hydrogen generation from radiolysis of the polymeric encapsulants, the met mounts are problematic for eventual disposal in a geologic repository

  6. Advanced Fuel Cycle Cost Basis

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2009-12-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  7. Advanced Fuel Cycle Cost Basis

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

    2007-04-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

  8. Advanced Fuel Cycle Cost Basis

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2008-03-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  9. The Snf1 Protein Kinase in the Yeast Saccharomyces cerevisiae

    Usaite, Renata

    2008-01-01

    4 on the regulation of glucose and galactose metabolism, I physiologically characterized Δsnf1, Δsnf4, and Δsnfsnf4 CEN.PK background yeast strains in glucose and glucose-galactose mixture batch cultivations (chapter 2). The results of this study showed that delayed induction of galactose...... that the stable isotope labeling approach is highly reproducible among biological replicates when complex protein mixtures containing small expression changes were analyzed. Where poor correlation between stable isotope labeling and spectral counting was found, the major reason behind the discrepancy was the lack...

  10. Effects of Fuel Quantity on Soot Formation Process for Biomass-Based Renewable Diesel Fuel Combustion

    Jing, Wei; Wu, Zengyang; Roberts, William L.; Fang, Tiegang

    2016-01-01

    Soot formation process was investigated for biomass-based renewable diesel fuel, such as biomass to liquid (BTL), and conventional diesel combustion under varied fuel quantities injected into a constant volume combustion chamber. Soot measurement

  11. Property-process relationships in nuclear fuel fabrication

    Tikare, V.

    2015-01-01

    Nuclear fuels are fabricated using many different techniques as they come in a large variety of shapes and compositions. The design and composition of nuclear fuels are predominantly dictated by the engineering requirements necessary for their function in reactors of various designs. Other engineering properties requirements originate from safety and security concerns, and the easy of handling, storing, transporting and disposing of the radioactive materials. In this chapter, the more common of these fuels will be briefly reviewed and the methods used to fabricate them will be presented. The fuels considered in this paper are oxide fuels used in LWRs and FRs, metal fuels in FRs and particulate fuels used in HTGRs. Fabrication of alternative fuel forms and use of standard fuels in alternative reactors will be discussed briefly. The primary motivation to advance fuel fabrication is to improve performance, reduce cost, reduce waste or enhance safety and security of the fuels. To achieve optimal performance, developing models to advance fuel fabrication has to be done in concert with developing fuel performance models. The specific properties and microstructures necessary for improved fuel performance must be identified using fuel performance models, while fuel fabrication models that can determine processing variables to give the desired microstructure and materials properties must be developed. (author)

  12. An innovative way of thinking nuclear waste management - Neutron physics of a reactor directly operating on SNF.

    Bruno Merk

    Full Text Available A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T promises a solution for improved waste management. Current strategies rely on systems designed in the 60's for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient.

  13. An innovative way of thinking nuclear waste management – Neutron physics of a reactor directly operating on SNF

    Litskevich, Dzianis; Bankhead, Mark; Taylor, Richard J.

    2017-01-01

    A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T) promises a solution for improved waste management. Current strategies rely on systems designed in the 60’s for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF) without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient. PMID:28749952

  14. Development of Advanced Spent Fuel Management Process

    Seo, Chung Seok; Choi, I. K.; Kwon, S. G. (and others)

    2007-06-15

    As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm{sup 2} and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields.

  15. Development of Advanced Spent Fuel Management Process

    Seo, Chung Seok; Choi, I. K.; Kwon, S. G.

    2007-06-01

    As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm 2 and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields

  16. Characterization plan for Hanford spent nuclear fuel

    Abrefah, J.; Thornton, T.A.; Thomas, L.E.; Berting, F.M.; Marschman, S.C.

    1994-12-01

    Reprocessing of spent nuclear fuel (SNF) at the Hanford Site Plutonium-Uranium Extraction Plant (PUREX) was terminated in 1972. Since that time a significant quantity of N Reactor and Single-Pass Reactor SNF has been stored in the 100 Area K-East (KE) and K-West (KW) reactor basins. Approximately 80% of all US Department of Energy (DOE)-owned SNF resides at Hanford, the largest portion of which is in the water-filled KE and KW reactor basins. The basins were not designed for long-term storage of the SNF and it has become a priority to move the SNF to a more suitable location. As part of the project plan, SNF inventories will be chemically and physically characterized to provide information that will be used to resolve safety and technical issues for development of an environmentally benign and efficient extended interim storage and final disposition strategy for this defense production-reactor SNF

  17. Licensing Air and Transboundary Shipments of Spent Nuclear Fuel

    Komarov, S.V.; Budu, M.E.; Derganov, D.V.; Savina, O.A.; Bolshinsky, I.M.; Moses, S.D.; Biro, L.

    2016-01-01

    Since 1996 the IAEA TS-R-1 regulation included new requirements applicable to transport of fissile materials by air. The later 2005 and 2009 editions confirmed the validity of those provisions. Despite the fact that the IAEA TS-R-1 allows for air shipments of SNF in Type B and Type C packages, the examples of such shipments are not abundant. Nuclear regulatory bodies and transport safety experts are cautious about air shipments of SNF. Why so? What are the risks? What are the alternatives? In this new regulatory framework, in 2009, two air shipments in Type B packages of Research Reactor (RR) Spent Nuclear Fuel (SNF) from Romania and Libya were performed under the U.S. DOE/NNSA RRRFR Program. The first licensing process of such shipment brought up many questions about package and shipment safety from the licensing experts' side and so the scope of analyses exceeded the requirements of IAEA. Under the thorough supervision of Rosatom and witnessed by DOE and CNCAN, all questions were answered by various strength analyses and risk evaluations. But the progress achieved didn't stop here. In 2010-2011, an energy absorption container (EAC) with titanium spheres as absorbers based on the SKODA VPVR/M cask was designed as the first Type C package in the world destined for RR SNF, currently under approval process. At the same time, intense preparations for the safe removal of the Russian-origin damaged RR SNF from Serbia, Vinca were in progress. The big amount of SNF and its rapidly worsening condition imposed as requirements to organize only one shipment as fast as possible, i.e. using at the maximum extent the entire experience available from other SNF shipments. The long route, several transit countries and means of transport, two different casks, new European regulations and many other issues resulted for the Serbian shipment in one of the most complex SNF shipments’ licensing exercise. This paper shows how the international regulatory framework ensures the

  18. Cost reductions of fuel cells for transport applications: fuel processing options

    Teagan, W P; Bentley, J; Barnett, B [Arthur D. Little, Inc., Cambridge, MA (United States)

    1998-03-15

    The highly favorable efficiency/environmental characteristics of fuel cell technologies have now been verified by virtue of recent and ongoing field experience. The key issue regarding the timing and extent of fuel cell commercialization is the ability to reduce costs to acceptable levels in both stationary and transport applications. It is increasingly recognized that the fuel processing subsystem can have a major impact on overall system costs, particularly as ongoing R and D efforts result in reduction of the basic cost structure of stacks which currently dominate system costs. The fuel processing subsystem for polymer electrolyte membrane fuel cell (PEMFC) technology, which is the focus of transport applications, includes the reformer, shift reactors, and means for CO reduction. In addition to low cost, transport applications require a fuel processor that is compact and can start rapidly. This paper describes the impact of factors such as fuel choice operating temperature, material selection, catalyst requirements, and controls on the cost of fuel processing systems. There are fuel processor technology paths which manufacturing cost analyses indicate are consistent with fuel processor subsystem costs of under $150/kW in stationary applications and $30/kW in transport applications. As such, the costs of mature fuel processing subsystem technologies should be consistent with their use in commercially viable fuel cell systems in both application categories. (orig.)

  19. Role of ion chromatograph in nuclear fuel fabrication process at Nuclear Fuel Complex

    Balaji Rao, Y.; Prasada Rao, G.; Prahlad, B.; Saibaba, N.

    2012-01-01

    The present paper discusses the different applications of ion chromatography followed in nuclear fuel fabrication process at Nuclear Fuel Complex. Some more applications of IC for characterization of nuclear materials and which are at different stages of method development at Control Laboratory, Nuclear Fuel Complex are also highlighted

  20. Spent nuclear fuel project cold vacuum drying facility process water conditioning system design description

    IRWIN, J.J.

    1998-01-01

    This document provides the System Design Description (SDD) for the Cold Vacuum Drying Facility (CVDF) Process Water Conditioning (PWC) System. The SDD was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998), the HNF-SD-SNF-DRD-O02, 1998, Cold Vacuum Drying Facility Design Requirements, and the CVDF Design Summary Report. The SDD contains general descriptions of the PWC equipment, the system functions, requirements and interfaces. The SDD provides references for design and fabrication details, operation sequences and maintenance. This SDD 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

  1. Method of processing spent fuel cladding tubes

    Nakatsuka, Masafumi; Ouchi, Atsuhiro; Imahashi, Hiromichi.

    1986-01-01

    Purpose: To decrease the residual activity of spent fuel cladding tubes in a short period of time and enable safety storage with simple storage equipments. Constitution: Spent fuel cladding tubes made of zirconium alloys discharged from a nuclear fuel reprocessing step are exposed to a grain boundary embrittling atmosphere to cause grain boundary destruction. This causes grain boundary fractures to the zirconium crystal grains as the matrix of nuclear fuels and then precipitation products precipitated to the grain boundary fractures are removed. The zirconium constituting the nuclear fuel cladding tube and other ingredient elements contained in the precipitation products are separated in this removing step and they are separately stored respectively. As a result, zirconium constituting most part of the composition of the spent nuclear fuel cladding tubes can be stored safely at a low activity level. (Takahashi, M.)

  2. Treatment and storage of high-level activity RAW and spent fuel from nuclear facilities

    Tomov, E.

    2010-01-01

    The most acceptable for the development of nuclear energy sector scenario is processing, storage and disposal of all SNF and waste from in the country of origin. Linking the supply of fresh nuclear fuel with subsequent transportation and processing would solve many of the problems related to its storage and accumulation at the site of the operator of the facility. Construction of NPP Belene is a prerequisite for a favorable solution to the management of SNF and HLW. At the stage of feasibility study for the construction of a deep geological repository, the studies of variants of the quantities of HLW from SNF reprocessing allow for a preliminary assessment of the capacity of the storage facility

  3. Nuclear fuel re-processing plant

    Sasaki, Yuko; Honda, Takashi; Shoji, Saburo; Kobayashi, Shiro; Furuya, Yasumasa

    1989-01-01

    In a nuclear fuel re-processing plant, high Si series stainless steels not always have sufficient corrosion resistance in a solution containing only nitric acid at medium or high concentration. Further, a method of blowing NOx gases may possibly promote the corrosion of equipment constituent materials remarkably. In view of the above, the corrosion promoting effect of nuclear fission products is suppressed without depositing corrosive metal ions as metals in the nitric acid solution. That is, a reducing atmosphere is formed by generating NOx by electrolytic reduction thereby preventing increase in the surface potential of stainless steels. Further, an anode is disposed in the nitric acid solution containing oxidative metal ions to establish an electrical conduction and separate them by way of partition membranes and a constant potential or constant current is applied while maintaining an ionic state so as not to deposit metals. Thus, equipments of re-processing facility can be protected from corrosion with no particular treatment for wastes as radioactive materials. (K.M.)

  4. The investigation of HTGR fuel regeneration process

    Lazarev, L N; Bertina, L E; Popik, V P; Isakov, V P; Alkhimov, N B; Pokhitonov, Yu A

    1985-07-01

    The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning.

  5. Electrochemical processing of spent nuclear fuel

    Williamson, M. A.; Willit, J. L.; Barnes, L. A.; Figueroa, J.; Limmer, S. L.; Blaskovitz, R. [Argonne National Laboratory, Argonne (United States)

    2008-08-15

    Our work in developing the fuel cycles and electrochemical technologies needed for the treatment of spent light water reactor and spent fast reactor fuel is progressing well. Baseline flowsheets along with a theoretical material balance have been developed for treatment of each type of fuel. A discussion about the flowsheets provides the opportunity to present the status of our technology development activities and future research and development directions.

  6. Electrochemical processing of spent nuclear fuel

    Williamson, M. A.; Willit, J. L.; Barnes, L. A.; Figueroa, J.; Limmer, S. L.; Blaskovitz, R.

    2008-01-01

    Our work in developing the fuel cycles and electrochemical technologies needed for the treatment of spent light water reactor and spent fast reactor fuel is progressing well. Baseline flowsheets along with a theoretical material balance have been developed for treatment of each type of fuel. A discussion about the flowsheets provides the opportunity to present the status of our technology development activities and future research and development directions

  7. The investigation of HTGR fuel regeneration process

    Lazarev, L.N.; Bertina, L.E.; Popik, V.P.; Isakov, V.P.; Alkhimov, N.B.; Pokhitonov, Yu.A.

    1985-01-01

    The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning

  8. Fuel Cell Stations Automate Processes, Catalyst Testing

    2010-01-01

    Glenn Research Center looks for ways to improve fuel cells, which are an important source of power for space missions, as well as the equipment used to test fuel cells. With Small Business Innovation Research (SBIR) awards from Glenn, Lynntech Inc., of College Station, Texas, addressed a major limitation of fuel cell testing equipment. Five years later, the company obtained a patent and provided the equipment to the commercial world. Now offered through TesSol Inc., of Battle Ground, Washington, the technology is used for fuel cell work, catalyst testing, sensor testing, gas blending, and other applications. It can be found at universities, national laboratories, and businesses around the world.

  9. Fuel fabrication and reprocessing for nuclear fuel cycle with inherent safety demands

    Shadrin, Andrey Yurevich; Dvoeglazov, Konstantin Nikolaevich; Ivanov, Valentine Borisovich; Volk, Vladimir Ivanovich; Skupov, Mikhail Vladimirovich; Glushenkov, Alexey Evgenevich [Joint Stock Company ' ' The High Technological Research Institute of Inorganic Materials' ' , Moscow (Russian Federation); Troyanov, Vladimir Mihaylovich; Zherebtsov, Alexander Anatolievich [Innovation and Technology Center of Project ' ' PRORYV' ' , State Atomic Energy Corporation ' ' Rosatom' ' , Moscow (Russian Federation)

    2015-06-01

    The strategies adopted in Russia for a closed nuclear fuel cycle with fast reactors (FR), selection of fuel type and recycling technologies of spent nuclear fuel (SNF) are discussed. It is shown that one of the possible technological solutions for the closing of a fuel cycle could be the combination of pyroelectrochemical and hydrometallurgical methods of recycling of SNF. This combined scheme allows: recycling of SNF from FR with high burn-up and short cooling time; decreasing the volume of stored SNF and the amount of plutonium in a closed fuel cycle in FR; recycling of any type of SNF from FR; obtaining the high pure end uranium-plutonium-neptunium end-product for fuel refabrication using pellet technology.

  10. Deployment Evaluation Methodology for the Electrometallurgical Treatment of DOE-EM Spent Nuclear Fuel

    Ramer, Ronald James; Adams, James Paul; Rynearson, Michael Ardel; Dahl, Christian Adam

    1999-01-01

    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). The NSNFP, conducted by Lockheed Martin Idaho Technology Co. (LMITCO) at the Idaho National Engineering and Environmental Laboratory (INEEL), is evaluating final disposition of SNF in the DOE complex. 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. Evaluations of treatment needs and options have been previously prepared, and further evaluations are ongoing activities in the DOE-EM NSNFP. 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. Treatment options and treatment locations will depend on fuel type and location of the fuel. One of the first steps associated with selecting one or more sites for treating SNF in the DOE complex is to determine the cost of each option. An economic analysis will assist in determining which fuel treatment alternative attains the optimum disposition of SNF at the lowest possible cost to the government and the public. One of the major issues associated with SNF treatment is final disposition of treatment products and associated waste streams. During conventional SNF treatment, various chemicals are added that may increase the product

  11. Mixed U/Pu oxide fuel fabrication facility co-processed feed, pelletized fuel

    1978-09-01

    Two conceptual MOX fuel fabrication facilities are discussed in this study. The first facility in the main body of the report is for the fabrication of LWR uranium dioxide - plutonium dioxide (MOX) fuel using co-processed feed. The second facility in the addendum is for the fabrication of co-processed MOX fuel spiked with 60 Co. Both facilities produce pellet fuel. The spiked facility uses the same basic fabrication process as the conventional MOX plant but the fuel feed incorporates a high energy gamma emitter as a safeguard measure against diversion; additional shielding is added to protect personnel from radiation exposure, all operations are automated and remote, and normal maintenance is performed remotely. The report describes the fuel fabrication process and plant layout including scrap and waste processing; and maintenance, ventilation and safety measures

  12. Modified ADS molten salt processes for back-end fuel cycle of PWR spent fuel

    Choi, In-Kyu; Yeon, Jei-Won; Kim, Won-Ho

    2002-01-01

    The back-end fuel cycle concept for PWR spent fuel is explained. This concept is adequate for Korea, which has operated both PWR and CANDU reactors. Molten salt processes for accelerator driven system (ADS) were modified both for the transmutation of long-lived radioisotopes and for the utilisation of the remained fissile uranium in PWR spent fuels. Prior to applying molten salt processes to PWR fuel, hydrofluorination and fluorination processes are applied to obtain uranium hexafluoride from the spent fuel pellet. It is converted to uranium dioxide and fabricated into CANDU fuel. From the remained fluoride compounds, transuranium elements can be separated by the molten salt technology such as electrowinning and reductive extraction processes for transmutation purpose without weakening the proliferation resistance of molten salt technology. The proposed fuel cycle concept using fluorination processes is thought to be adequate for our nuclear program and can replace DUPIC (Direct Use of spent PWR fuel in CANDU reactor) fuel cycle. Each process for the proposed fuel cycle concept was evaluated in detail

  13. Progress on the Hanford K basins spent nuclear fuel project

    Culley, G.E.; Fulton, J.C.; Gerber, E.W.

    1996-01-01

    This paper highlights progress made during the last year toward removing the Department of Energy's (DOE) approximately, 2,100 metric tons of metallic spent nuclear fuel from the two outdated K Basins at the Hanford Site and placing it in safe, economical interim dry storage. In the past year, the Spent Nuclear Fuel (SNF) Project has engaged in an evolutionary process involving the customer, regulatory bodies, and the public that has resulted in a quicker, cheaper, and safer strategy for accomplishing that goal. Development and implementation of the Integrated Process Strategy for K Basins Fuel is as much a case study of modern project and business management within the regulatory system as it is a technical achievement. A year ago, the SNF Project developed the K Basins Path Forward that, beginning in December 1998, would move the spent nuclear fuel currently stored in the K Basins to a new Staging and Storage Facility by December 2000. The second stage of this $960 million two-stage plan would complete the project by conditioning the metallic fuel and placing it in interim dry storage by 2006. In accepting this plan, the DOE established goals that the fuel removal schedule be accelerated by a year, that fuel conditioning be closely coupled with fuel removal, and that the cost be reduced by at least $300 million. The SNF Project conducted coordinated engineering and technology studies over a three-month period that established the technical framework needed to design and construct facilities, and implement processes compatible with these goals. The result was the Integrated Process Strategy for K Basins Fuel. This strategy accomplishes the goals set forth by the DOE by beginning fuel removal a year earlier in December 1997, completing it by December 1999, beginning conditioning within six months of starting fuel removal, and accomplishes it for $340 million less than the previous Path Forward plan

  14. Assessment results of the South Korea TRIGA SNF to be shipped to INEEL

    Cole, C.M.; Dirk, W.J.; Cottam, R.E.; Paik, S.T.

    1997-01-01

    This paper describes the Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) examination at the Seoul and the Taejon Research Reactor Facilities in South Korea. The examination was required before the SNF would be accepted for transportation and storage at the INEEL. The results of the aluminum and stainless steel clad TRIGA fuel examination have been summarized. A description of the examination team training, the examination work plan and examination equipment is also included. This paper also explains the technical basis for the examination and physical condition criteria used to determine what, if any, additional packaging would be required for transportation and for the receipt and storage of the fuel at the INEEL. This paper delineates the preparation activities prior to the fuel examinations and includes (1) collecting spent fuel data; (2) preparatory work by the Korean Atomic Energy Research Institute (KAERI) for fuel examination: (3) preparation of a radionuclide report, Radionuclide Mass Inventory, Activity, Decay Heat, and Dose Rate Parametric Data for TRIGA Spent Nuclear Fuels needed to provide input data for transportation and fuel acceptance at the Idaho National Engineering and Environmental Laboratory (INEEL); (4) gathering FRR Facility data; and (5) coordination between the INEEL and KAERI. Included, are the unanticipated conditions encountered in the unloading of fuel from the dry storage casks in Taejon in preparation for examination, a description of the damaged condition of the fuel removed from the casks, and the apparent cause of the damages. Lessons learned from all the activities are also addressed. A brief description of the preparatory work for the shipment of the spent fuel from Korea to INEEL is included

  15. Assessment results of the South Korea TRIGA SNF to be shipped to INEEL

    Cole, Charles M.; Dirk, Willam J.; Cottam, Russel E.; Paik, Sam T.

    1997-01-01

    This paper describes the Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) examination at the Seoul and the Taejon Research Reactor Facilities in South Korea. The examination was required before the SNF would be accepted for transportation and storage at the INEEL. The results of the aluminum and stainless steel clad TRIGA fuel examination have been summarized. A description of the examination team training, the examination work plan and examination equipment is also included. This paper also explains the technical basis for the examination and physical condition criteria used to determine what, if any, additional packaging (canning) would be required for transportation and for the receipt and storage of the fuel at the INEEL. This paper delineates the preparation activities prior to the fuel examinations and includes (1) collecting spent fuel data; (2) preparatory work by the Korean Atomic Energy Research Institute (KAERI) for fuel examination: (3) preparation of a radionuclide report, 'Radionuclide Mass Inventory, Activity, Decay Heat, and Dose Rate Parametric Data for TRIGA Spent Nuclear Fuels' needed to provide input data for transportation and fuel acceptance at the Idaho National Engineering and Environmental Laboratory (INEEL); (4) gathering FRR Facility data; (5) preparation of Appendix A; (6) and coordination between the INEEL and KAERI. Included, are the unanticipated conditions encountered in the unloading of fuel from the dry storage casks in Taejon in preparation for examination, a description of the damaged condition of the fuel removed from the casks, and the apparent cause of the damages. Lessons learned from all the activities are also addressed. A brief description of the preparatory work for the shipment of the spent fuel from Korea to INEEL is included. (author)

  16. Materials and processes for solar fuel production

    Viswanathan, Balasubramanian; Lee, Jae Sung

    2014-01-01

    This book features different approaches to non-biochemical pathways for solar fuel production. This one-of-a-kind book addresses photovoltaics, photocatalytic water splitting for clean hydrogen production and CO2 conversion to hydrocarbon fuel through in-depth comprehensive contributions from a select blend of established and experienced authors from across the world. The commercial application of solar based systems, with particular emphasis on non-PV based devices have been discussed. This book intends to serve as a primary resource for a multidisciplinary audience including chemists, engineers and scientists providing a one-stop location for all aspects related to solar fuel production. The material is divided into three sections: Solar assisted water splitting to produce hydrogen; Solar assisted CO2 utilization to produce green fuels and Solar assisted electricity generation. The content strikes a balance between theory, material synthesis and application with the central theme being solar fuels.

  17. Handbook on process and chemistry on nuclear fuel reprocessing

    Suzuki, Atsuyuki [Tokyo Univ., Tokyo (Japan); Asakura, Toshihide; Adachi, Takeo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others

    2001-12-01

    'Wet-type' nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of 'wet-type' reprocessing, because it contributes to establish and develop fuel reprocessing process and nuclear fuel cycle treating high burn-up UO{sub 2} fuel and spent MOX fuel, and to utilize 'wet-type' reprocessing technology much widely. This handbook summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing', from FY 1993 until FY 2000. (author)

  18. Handbook on process and chemistry on nuclear fuel reprocessing

    Suzuki, Atsuyuki; Asakura, Toshihide; Adachi, Takeo

    2001-12-01

    'Wet-type' nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of 'wet-type' reprocessing, because it contributes to establish and develop fuel reprocessing process and nuclear fuel cycle treating high burn-up UO 2 fuel and spent MOX fuel, and to utilize 'wet-type' reprocessing technology much widely. This handbook summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing', from FY 1993 until FY 2000. (author)

  19. Handbook on process and chemistry on nuclear fuel reprocessing

    Suzuki, Atsuyuki (ed.) [Tokyo Univ., Tokyo (Japan); Asakura, Toshihide; Adachi, Takeo (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2001-12-01

    'Wet-type' nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of 'wet-type' reprocessing, because it contributes to establish and develop fuel reprocessing process and nuclear fuel cycle treating high burn-up UO{sub 2} fuel and spent MOX fuel, and to utilize 'wet-type' reprocessing technology much widely. This handbook summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing', from FY 1993 until FY 2000. (author)

  20. Analysis of Ignition Testing on K-West Basin Fuel

    J. Abrefah; F.H. Huang; W.M. Gerry; W.J. Gray; S.C. Marschman; T.A. Thornton

    1999-08-10

    Approximately 2100 metric tons of spent nuclear fuel (SNF) discharged from the N-Reactor have been stored underwater at the K-Basins in the 100 Area of the Hanford Site. The spent fuel has been stored in the K-East Basin since 1975 and in the K-West Basin since 1981. Some of the SNF elements in these basins have corroded because of various breaches in the Zircaloy cladding that occurred during fuel discharge operations and/or subsequent handling and storage in the basins. Consequently, radioactive material in the fuel has been released into the basin water, and water has leaked from the K-East Basin into the soil below. To protect the Columbia River, which is only 380 m from the basins, the SNF is scheduled to be removed and transported for interim dry storage in the 200 East Area, in the central portion of the Site. However, before being shipped, the corroded fuel elements will be loaded into Multi-Canister OverPacks and conditioned. The conditioning process will be selected based on the Integrated Process Strategy (IPS) (WHC 1995), which was prepared on the basis of the dry storage concept developed by the Independent Technical Assessment (ITA) team (ITA 1994).

  1. Analysis of Ignition Testing on K-West Basin Fuel

    Abrefah, J.; Huang, F.H.; Gerry, W.M.; Gray, W.J.; Marschman, S.C.; Thornton, T.A.

    1999-01-01

    Approximately 2100 metric tons of spent nuclear fuel (SNF) discharged from the N-Reactor have been stored underwater at the K-Basins in the 100 Area of the Hanford Site. The spent fuel has been stored in the K-East Basin since 1975 and in the K-West Basin since 1981. Some of the SNF elements in these basins have corroded because of various breaches in the Zircaloy cladding that occurred during fuel discharge operations and/or subsequent handling and storage in the basins. Consequently, radioactive material in the fuel has been released into the basin water, and water has leaked from the K-East Basin into the soil below. To protect the Columbia River, which is only 380 m from the basins, the SNF is scheduled to be removed and transported for interim dry storage in the 200 East Area, in the central portion of the Site. However, before being shipped, the corroded fuel elements will be loaded into Multi-Canister OverPacks and conditioned. The conditioning process will be selected based on the Integrated Process Strategy (IPS) (WHC 1995), which was prepared on the basis of the dry storage concept developed by the Independent Technical Assessment (ITA) team (ITA 1994)

  2. Evaluation of helium impurity impacts on Spent Nuclear Fuel project processes (OCRWM)

    SHERRELL, D.L.

    1999-09-21

    This document identifies the types and quantities of impurities that may be present within helium that is introduced into multi-canister overpacks (MCO)s by various SNF Project facilities, including, but not limited to the Cold Vacuum Drying (CVD) Facility (CVDF). It then evaluates possible impacts of worst case impurity inventories on MCO drying, transportation, and storage processes. Based on the evaluation results, this document: (1) concludes that the SNF Project helium procurement specification can be a factor-of-ten less restrictive than a typical vendor's standard offering (99.96% pure versus the vendor's 99.997% pure standard offering); (2) concludes that the CVDF's current 99.5% purity requirement is adequate to control the quality of the helium that is delivered to the MCO by the plant's helium distribution system; and (3) recommends specific impurity limits for both of the above cases.

  3. Evaluation of helium impurity impacts on Spent Nuclear Fuel project processes (OCRWM)

    SHERRELL, D.L.

    1999-01-01

    This document identifies the types and quantities of impurities that may be present within helium that is introduced into multi-canister overpacks (MCO)s by various SNF Project facilities, including, but not limited to the Cold Vacuum Drying (CVD) Facility (CVDF). It then evaluates possible impacts of worst case impurity inventories on MCO drying, transportation, and storage processes. Based on the evaluation results, this document: (1) concludes that the SNF Project helium procurement specification can be a factor-of-ten less restrictive than a typical vendor's standard offering (99.96% pure versus the vendor's 99.997% pure standard offering); (2) concludes that the CVDF's current 99.5% purity requirement is adequate to control the quality of the helium that is delivered to the MCO by the plant's helium distribution system; and (3) recommends specific impurity limits for both of the above cases

  4. Recent advances in fuel product and manufacturing process development

    Slember, R.J.; Doshi, P.K.

    1987-01-01

    This paper discusses advancements in commercial nuclear fuel products and manufacturing made by the Westinghouse Electric Corporation in response to the commercial nuclear fuel industry's demand for high reliability, increased plant availability and improved operating flexibility. The features and benefits of Westinghouse's most advanced fuel products--VANTAGE 5 for PWR plants and QUAD+ for BWR plants--are described, as well as 'high performance' fuel concepts now under development for delivery in the late 1980s. The paper also disusses the importance of in-process quality control throughout manufacturing towards reducing product variability and improving fuel reliability. (author)

  5. Interface agreement for the management of FFTF Spent Nuclear Fuel

    McCormack, R.L.

    1995-01-01

    The Hanford Site Spent Nuclear Fuel (SNF) Project was formed to manage the SNF at Hanford. The mission of the Fast Flux Test Facility (FFTF) Transition Project is to place the facility in a radiologically and industrially safe shutdown condition for turnover to the Environmental Restoration Contractor (ERC) for subsequent D ampersand D. To satisfy both project missions, FFTF SNF must be removed from the FFTF and subsequently dispositioned. This documented provides the interface agreement between FFTF Transition Project and SNF Project for management of the FFTF SNF

  6. Process for humidifying a gaseous fuel stream

    Sederquist, R. A.

    1985-01-01

    A fuel gas stream for a fuel cell is humidified by a recirculating hot liquid water stream using the heat of condensation from the humidified stream as the heat to vaporize the liquid water. Humidification is accomplished by directly contacting the liquid water with the dry gas stream in a saturator to evaporate a small portion of water. The recirculating liquid water is reheated by direct contact with the humidified gas stream in a condenser, wherein water is condensed into the liquid stream. Between the steps of humidifying and condensing water from the gas stream it passes through the fuel cell and additional water, in the form of steam, is added thereto

  7. DUPIC nuclear fuel manufacturing and process technology development

    Yang, Myung Seung; Park, J. J.; Lee, J. W.

    2000-05-01

    In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated

  8. DUPIC nuclear fuel manufacturing and process technology development

    Yang, Myung Seung; Park, J. J.; Lee, J. W. [and others

    2000-05-01

    In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated.

  9. A Post Closure Safety Assessment for Radioactive Wastes from Advanced nuclear fuel Cycle

    Kang, Chul Hyung; Hwang, Yong Soo

    2010-01-01

    KAERI has developed the KIEP-21 (Korean, Innovative, Environmentally Friendly, and Proliferation Resistant System for the 21st Century). It is an advanced nuclear fuel cycle option with a pyro-process and a GEN-IV SFR. A pyro-process consists of two distinctive processes, an electrolytic reduction process and an electro-refining and winning process. When the pyro-process is applied, it generates five streams of wastes. To compare pyro-process advantage over the direct disposal of Spent Nuclear Fuel (SNF), the PWR SNF of the 45,000 MWD burn-up has been assumed. A safety assessment model for pyro-process wastes and representative results are presented in this report

  10. An Automated Process for Generation of New Fuel Breakdown Mechanisms

    Violi, Angela

    2006-01-01

    .... It combines advanced computational techniques in a synergistic study of the critical processes in fuel decomposition at a level of detail that can help distinguish, correct, and quantify mechanisms for these processes...

  11. Fuel processing for molten-salt reactors

    Hightower, J.R. Jr.

    1975-01-01

    Progress is reported on the development of processes for the isolation of protactinium and for the removal of fission products from molten-salt breeder reactors. The metal transfer experiment MTE-3 (for removing rare earths from MSRE fuel salt) was completed and the equipment used in that experiment was examined. The examination showed that no serious corrosion had occurred on the internal surfaces of the vessels, but that serious air oxidation occurred on the external surfaces of the vessels. Analyses of the bismuth phases indicated that the surfaces in contact with the salts were enriched in thorium and iron. Mass transfer coefficients in the mechanically agitated nondispersing contactors were measured in the Salt/Bismuth Flow-through Facility. The measured mass transfer coefficients are about 30 to 40 percent of those predicted by the preferred literature correlation, but were not as low as those seen in some of the runs in MTE-3. Additional studies using water--mercury systems to simulate molten salt-bismuth systems indicated that the model used to interpret results from previous measurements in the water--mercury system has significant deficiencies. Autoresistance heating studies were continued to develop a means of internal heat generation for frozen-wall fluorinators. Equipment was built to test a design of a side arm for the heating electrode. Results of experiments with this equipment indicate that for proper operation the wall temperature must be held much lower than that for which the equipment was designed. Studies with an electrical analog of the equipment indicate that no regions of abnormally high current density exist in the side arm. (JGB)

  12. Deployment Evaluation Methodology for the Electrometallurgical Treatment of DOE-EM Spent Nuclear Fuel

    Dahl, C.A.; Adams, J.P.; Rynearson, M.A.; Ramer, R.J.

    1999-01-01

    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

  13. Modeling of large-scale oxy-fuel combustion processes

    Yin, Chungen

    2012-01-01

    Quite some studies have been conducted in order to implement oxy-fuel combustion with flue gas recycle in conventional utility boilers as an effective effort of carbon capture and storage. However, combustion under oxy-fuel conditions is significantly different from conventional air-fuel firing......, among which radiative heat transfer under oxy-fuel conditions is one of the fundamental issues. This paper demonstrates the nongray-gas effects in modeling of large-scale oxy-fuel combustion processes. Oxy-fuel combustion of natural gas in a 609MW utility boiler is numerically studied, in which...... calculation of the oxy-fuel WSGGM remarkably over-predicts the radiative heat transfer to the furnace walls and under-predicts the gas temperature at the furnace exit plane, which also result in a higher incomplete combustion in the gray calculation. Moreover, the gray and non-gray calculations of the same...

  14. Fuel transfer system ALARA design review - Project A.15

    KUEBERTH, L.R.

    2001-01-01

    One mission of the Spent Nuclear Fuel (SNF) Project is to move the SNF from the K Basins in the Hanford 100K Area to an interim dry storage at the Canister Storage Building (CSB) in the Hanford 200 East Area. The Fuel Transfer System (FTS) is a subproject that will move the SNF from the 105K East (KE) Facility to the 105K West (KW) Facility. The SNF will be treated for shipment to the Cold Vacuum Drying (CVD) facility at the KW Basin. The SNF canisters will be loaded underwater into a Shielded Transfer Cask (STC) in the KE Basin. The fully loaded STC will be brought out of the water and placed into a Cask Transfer Overpack (CTO) by the STC Straddle Carrier. As the STC is removed from the water, it will be washed down with demineralized water by an manual rinse system. The CTO with the STC inside will be placed on a transport trailer and transferred to the KW Basin as an intra-facility transfer. The CTO will be unloaded from the shipping trailer at the KW Basin and the STC will be removed from the CTO. The STC will then be lowered into the KW Basin water and the fuel will be removed. The SNF will then be processed for shipment to the CVD. As soon as all of the fuel has been removed from the STC, the cask will be removed from the KW Basin water and placed into the CTO. The CTO will again be placed on the trailer for transport back to the KE Basin where the entire cycle will be repeated approximately 400 times. This document records the As Low As Reasonably Achievable (ALARA) findings and design recommendations/requirements by the SNF Project noted during the Final Design Review of the STC, CTO, STC Transfer System, Annexes and Roadways for support of FTS. This document is structured so that all statements that include the word ''shall'' represent design features that have been or will be implemented within the project scope. Statements that include the words ''should'' or ''recommend'' represent ALARA design features to be evaluated for future implementation

  15. Modeling the Thermal Rocket Fuel Preparation Processes in the Launch Complex Fueling System

    A. V. Zolin

    2015-01-01

    Full Text Available It is necessary to carry out fuel temperature preparation for space launch vehicles using hydrocarbon propellant components. A required temperature is reached with cooling or heating hydrocarbon fuel in ground facilities fuel storages. Fuel temperature preparing processes are among the most energy-intensive and lengthy processes that require the optimal technologies and regimes of cooling (heating fuel, which can be defined using the simulation of heat exchange processes for preparing the rocket fuel.The issues of research of different technologies and simulation of cooling processes of rocket fuel with liquid nitrogen are given in [1-10]. Diagrams of temperature preparation of hydrocarbon fuel, mathematical models and characteristics of cooling fuel with its direct contact with liquid nitrogen dispersed are considered, using the numerical solution of a system of heat transfer equations, in publications [3,9].Analytical models, allowing to determine the necessary flow rate and the mass of liquid nitrogen and the cooling (heating time fuel in specific conditions and requirements, are preferred for determining design and operational characteristics of the hydrocarbon fuel cooling system.A mathematical model of the temperature preparation processes is developed. Considered characteristics of these processes are based on the analytical solutions of the equations of heat transfer and allow to define operating parameters of temperature preparation of hydrocarbon fuel in the design and operation of the filling system of launch vehicles.The paper considers a technological system to fill the launch vehicles providing the temperature preparation of hydrocarbon gases at the launch site. In this system cooling the fuel in the storage tank before filling the launch vehicle is provided by hydrocarbon fuel bubbling with liquid nitrogen. Hydrocarbon fuel is heated with a pumping station, which provides fuel circulation through the heat exchanger-heater, with

  16. Experience Of Using Metal-and-Concrete Cask TUK-108/1 For Storage And Transportation Of Spent Nuclear Fuel Of Decommissioned NPS

    Barnes, E.; Dyer, R. [Environmental Protection Agency, Ronald Reagan Bldg. 3rd Floor 1200 Pennsylvania Av., NW Washington, D.C. 20024 (United States); Snipes, R. [Oak Ridge National Laboratories, VA (United States); Dolbenkov, V.G.; Guskov, V.D.; Korotkov, G.V. [Joint Stock Company ' KBSM' , 64 Lesnoy Av., St.Petersburg 194100 (Russian Federation); Makarchuk, T.F. [Joint Stock Company ' Atomstroyexport' , Potapovskiy str. 5, bld. 4, Moscow, 101990 (Russian Federation); Zakharchev, A.A. [State Corporation ' Rosatom' , 24-26 Ordinka St., Moscow, 100000 (Russian Federation)

    2009-06-15

    In past 10 years in Russia an intensive development of a new technology of management of spent nuclear fuel (SNF) has taken place. This technology is based on the concept of using a shielded cask which provides safety of its content (SNF) and meeting all other safety requirements to storage and transportation of SNF. Radiation protection against emission and non-propagation of activity outside the cask is ensured by the physical barriers such as all-metal or composite body, face work, inner structures to accommodate spent fuel assemblies (SFA), lids with sealing systems. Residual heat buildup is off-taken to the environment by natural way: emission and convection of surrounding air. The necessity in development of the cask technology of SNF management was conditioned by the situation at hand with defueling of Russian decommissioned nuclear-powered submarines (NPS) as the existed transport infrastructure and enterprises involved in fuel processing could not meet the demand for transportation and processing of SNF neither from reactors of all dismantled NPS, nor from reactors of NPS waiting for decommissioning. The US and Norway actively participated in the trilateral joint project with the Russian Federation aimed at creation of a cask prototype for interim storage and transportation of SNF of dismantled NPS. The 1.1 Project is a part of the Arctic Military Environmental Cooperation (AMEC) Program. In December 2000 the project was successfully completed by issuance of the certificate-permit for design and transportation of NP Submarine SNF. It was a first certified dual-purpose TUK from the MMC family. In these years 106 TUK-108/1 casks have been manufactured and supplied to PO Mayak, JSC CS Zvezda, JSC CS Zvezdochka and FSUE DalRAO. The storage pads for interim storage of TUK-108/1 have been built and currently are in operation on sites of SNF unloading from submarine reactors and SNF cask-loading such as JSC CS Zvezda, JSC CS Zvezdochka and FSUE DalRAO. In

  17. INEL integrated spent nuclear fuel consolidation task team report

    Henry, R.N.; Clark, J.H.; Chipman, N.A.

    1994-01-01

    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

  18. PWR-to-PWR fuel cycle model using dry process

    Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

    2002-03-01

    PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

  19. Spent Nuclear Fuel Project document control and Records Management Program Description

    MARTIN, B.M.

    2000-01-01

    The Spent Nuclear Fuel (SNF) Project document control and records management program, as defined within this document, is based on a broad spectrum of regulatory requirements, Department of Energy (DOE) and Project Hanford and SNF Project-specific direction and guidance. The SNF Project Execution Plan, HNF-3552, requires the control of documents and management of records under the auspices of configuration control, conduct of operations, training, quality assurance, work control, records management, data management, engineering and design control, operational readiness review, and project management and turnover. Implementation of the controls, systems, and processes necessary to ensure compliance with applicable requirements is facilitated through plans, directives, and procedures within the Project Hanford Management System (PHMS) and the SNF Project internal technical and administrative procedures systems. The documents cited within this document are those which directly establish or define the SNF Project document control and records management program. There are many peripheral documents that establish requirements and provide direction pertinent to managing specific types of documents that, for the sake of brevity and clarity, are not cited within this document

  20. Management of spent nuclear fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee: Environmental assessment

    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

  1. Alteration in fuel processing at Tokai Works of Mitsubishi Nuclear Fuel Co., Ltd

    1977-01-01

    The report of the Committee on Examination of Nuclear Fuel Safety to the Atomic Energy Commission of Japan concerning the alteration is given, which is attached to the reply from the commission to the prime minister, and its safety was confirmed. The alterations are installation of the storage for transport containers containing fuel assemblies, construction of radiation control and other buildings; and improvement and installation of the facilities for chemical-processing, pellet fabrication, fuel assembling, and storage. (Mori, K.)

  2. Process for assembling a nuclear fuel element

    Wachtendonk, H.J. von.

    1984-01-01

    Before insertion into the spacers, the fuel rocks are coated with a self-hardening layer of water-soluble polyvinyl and/or polyether polymer to prevent scratches on the cladding tubes. After insertion, the protective conting is removed by means of water. (orig.) [de

  3. The pathway by which the yeast protein kinase Snf1p controls acquisition of sodium tolerance is different from that mediating glucose regulation.

    Ye, Tian; Elbing, Karin; Hohmann, Stefan

    2008-09-01

    It recently became apparent that the highly conserved Snf1p protein kinase plays roles in controlling different cellular processes in the yeast Saccharomyces cerevisiae, in addition to its well-known function in glucose repression/derepression. We have previously reported that Snf1p together with Gis4p controls ion homeostasis by regulating expression of ENA1, which encodes the Ena1p Na(+) extrusion system. In this study we found that Snf1p is rapidly phosphorylated when cells are exposed to NaCl and this phosphorylation is required for the role of Snf1p in Na(+) tolerance. In contrast to activation by low glucose levels, the salt-induced phosphorylation of Snf1p promoted neither phosphorylation nor nuclear export of the Mig1p repressor. The mechanism that prevents Mig1p phosphorylation by active Snf1p under salt stress does not involve either hexokinase PII or the Gis4p regulator. Instead, Snf1p may mediate upregulation of ENA1 expression via the repressor Nrg1p. Activation of Snf1p in response to glucose depletion requires any of the three upstream protein kinases Sak1p, Tos3p and Elm1p, with Sak1p playing the most prominent role. The same upstream kinases were required for salt-induced Snf1p phosphorylation, and also under these conditions Sak1p played the most prominent role. Unexpectedly, however, it appears that Elm1p plays a dual role in acquisition of salt tolerance by activating Snf1p and in a presently unknown parallel pathway. Together, these results indicate that under salt stress Snf1p takes part in a different pathway from that during glucose depletion and this role is performed together as well as in parallel with its upstream kinase Elm1p. Snf1p appears to be part of a wider functional network than previously anticipated and the full complexity of this network remains to be elucidated.

  4. State of the art of UO2 fuel fabrication processes

    Henke, M.; Klemm, U.

    1980-01-01

    Starting from the need of UO 2 for thermal power reactors in the period from 1980 to 1990 and the role of UF 6 conversion into UO 2 within the fuel cycle, the state-of-the-art of the three established industrial processes - ADU process, AUC process, IDR process - is assessed. The number of process stages and requirements on process management are discussed. In particular, the properties of the fabricated UO 2 powders, their influence on the following pellet production and on operational behaviour of the fuel elements under reactor conditions are described. Hence, an evaluation of the three essential conversion processes is derived. (author)

  5. Aspen Forest Cover by Stratum/Plot (SNF)

    National Aeronautics and Space Administration — Average percent coverage and standard deviation of each canopy stratum from subplots at each aspen site during the SNF study in the Superior National Forest, Minnesota

  6. Processing of spent nuclear fuel from light water reactors

    Sraier, V.

    1978-11-01

    A comprehensive review is given of the reprocessing of spent nuclear fuel from LWR's (covering references up to No. 18 (1977) of INIS inclusively). Particular attention is devoted to waste processing, safety, and reprocessing plants. In the addendum, the present status is shown on the example of KEWA, the projected large German fuel reprocessing plant. (author)

  7. Process for automatic filling of nuclear fuel rod cans

    Bezold, H.

    1977-01-01

    A drying section is inserted in the production line for the automation of the filling process for fuel rods with nuclear fuel pellets. The pellets are taken in a drum magazine to a drying furnace and then pushed out one after the other into the can to be filled. (TK) [de

  8. Purification and characterization of the three Snf1-activating kinases of Saccharomyces cerevisiae

    Elbing, Karin; McCartney, Rhonda R.; Schmidt, Martin C.

    2006-01-01

    Members of the Snf1/AMPK family of protein kinases are activated by distinct upstream kinases that phosphorylate a conserved threonine residue in the Snf1/AMPK activation loop. Recently, the identities of the Snf1- and AMPK-activating kinases have been determined. Here we describe the purification and characterization of the three Snf1-activating kinases of Saccharomyces cerevisiae. The identities of proteins associated with the Snf1-activating kinases were determined by peptide mass fingerpr...

  9. Synergistic energy conversion process using nuclear energy and fossil fuels

    Hori, Masao

    2007-01-01

    Because primary energies such as fossil fuels, nuclear energy and renewable energy are limited in quantity of supply, it is necessary to use available energies effectively for the increase of energy demand that is inevitable this century while keeping environment in good condition. For this purpose, an efficient synergistic energy conversion process using nuclear energy and fossil fuels together converted to energy carriers such are electricity, hydrogen, and synthetic fuels seems to be effective. Synergistic energy conversion processes containing nuclear energy were surveyed and effects of these processes on resource saving and the CO 2 emission reduction were discussed. (T.T.)

  10. Radioactive waste management of experimental DUPIC fuel fabrication process

    Lee, H. H.; Park, J. J.; Shin, J. M.; Yang, M. S.; Hong, K. P.

    2001-01-01

    The concept of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) is a dry processing technology to manufacture CANDU compatible DUPIC fuel from spent PWR fuel material. Real spent PWR fuel was used in IMEF M6 hot cell to carry out DUPIC experiment. Afterwards, about 200 kg-U of spent PWR fuel is supposed to be used till 2006. This study has been conducted in some hot cells of PIEF and M6 cell of IMEF. There are various forms of nuclear material such as rod cut, powder, green pellet, sintered pellet, fabrication debris, fuel rod, fuel bundle, sample, and process waste produced from various manufacturing experiment of DUPIC fuel. After completing test, the above nuclear wastes and test equipment etc. will be classified as radioactive waste, transferred to storage facility and managed rigorously according to domestic and international laws until the final management policy is determined. It is desirable to review management options in advance for radioactive waste generated from manufacturing experiment of DUPIC nuclear fuel as well as residual nuclear material and dismantled equipment. This paper includes basic plan for DUPIC radwaste, arising source and estimated amount of radioactive waste, waste classification and packing, transport cask, transport procedures

  11. Method for pre-processing LWR spent fuel

    Otsuka, Katsuyuki; Ebihara, Hikoe.

    1986-01-01

    Purpose: To facilitate the decladding of spent fuel, cladding tube processing, and waste gas recovery, and to enable the efficient execution of main re-processing process thereafter. Constitution: Spent fuel assemblies are sent to a cutting process where they are cut into chips of easy-to-process size. The chips, in a thermal decladding process, undergo a thermal cycle processing in air with the processing temperatures increased and decreased within the range of from 700 deg C to 1200 deg C, oxidizing zircaloy comprising the cladding tubes into zirconia. The oxidized cladding tubes have a number of fine cracks and become very brittle and easy to loosen off from fuel pellets when even a slight mechanical force is applied thereto, thus changing into a form of powder. Processed products are then separated into zirconia sand and fuel pellets by a gravitational selection method or by a sifting method, the zirconia sand being sent to a waste processing process and the fuel pellets to a melting-refining process. (Yoshino, Y.)

  12. Safety analysis of IFR fuel processing in the Argonne National Laboratory Fuel Cycle Facility

    Charak, I; Pedersen, D.R.; Forrester, R.J.; Phipps, R.D.

    1993-01-01

    The Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory (ANL) includes on-site processing and recycling of discharged core and blanket fuel materials. The process is being demonstrated in the Fuel Cycle Facility (FCF) at ANL's Idaho site. This paper describes the safety analyses that were performed in support of the FCF program; the resulting safety analysis report was the vehicle used to secure authorization to operate the facility and carry out the program, which is now under way. This work also provided some insights into safety-related issues of a commercial IFR fuel processing facility. These are also discussed

  13. Storage and treatment of SNF of Alfa class nuclear submarines: current status and problems

    Ignatiev, Sviatoslav; Zabudko, Alexey; Pankratov, Dmitry; Somov, Ivan; Suvorov, Gennady

    2007-01-01

    Available in abstract form only. Full text of publication follows: The current status and main problems associated with storage, defueling and following treatment of spent nuclear fuel (SNF) of Nuclear Submarines (NS) with heavy liquid metal cooled reactors are considered. In the final analysis these solutions could be realized in the form of separate projects to be funded through national and bi- and multilateral funding in the framework of the international collaboration of the Russian Federation on complex utilization of NS and rehabilitation of contaminated objects allocated in the North-West region of Russia. (authors)

  14. Effects of Fuel Quantity on Soot Formation Process for Biomass-Based Renewable Diesel Fuel Combustion

    Jing, Wei

    2016-12-01

    Soot formation process was investigated for biomass-based renewable diesel fuel, such as biomass to liquid (BTL), and conventional diesel combustion under varied fuel quantities injected into a constant volume combustion chamber. Soot measurement was implemented by two-color pyrometry under quiescent type diesel engine conditions (1000 K and 21% O2 concentration). Different fuel quantities, which correspond to different injection widths from 0.5 ms to 2 ms under constant injection pressure (1000 bar), were used to simulate different loads in engines. For a given fuel, soot temperature and KL factor show a different trend at initial stage for different fuel quantities, where a higher soot temperature can be found in a small fuel quantity case but a higher KL factor is observed in a large fuel quantity case generally. Another difference occurs at the end of combustion due to the termination of fuel injection. Additionally, BTL flame has a lower soot temperature, especially under a larger fuel quantity (2 ms injection width). Meanwhile, average soot level is lower for BTL flame, especially under a lower fuel quantity (0.5 ms injection width). BTL shows an overall low sooting behavior with low soot temperature compared to diesel, however, trade-off between soot level and soot temperature needs to be carefully selected when different loads are used.

  15. Gas turbine with two circuits and intermediate fuel conversion process

    Bachl, H.

    1978-01-01

    The combination of a fuel conversion process with a thermal process saves coolant and subsequent separation plant, in order to achieve the greatest possible use of the mechanical or electrical energy. The waste heat of a thermal circuit is taken to an endothermal chemical fuel conversion process arranged before a second circuit. The heat remaining after removal of the heat required for the chemical process is taken to a second thermal circuit. The reaction products of the chemical process which condense out during expansion in the second thermal process are selectively separated from the remaining gas mixture in the individual turbine stages. (HGOE) [de

  16. Managing Spent Nuclear Fuel at the Idaho National Laboratory

    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.

  17. Systems and processes for conversion of ethylene feedstocks to hydrocarbon fuels

    Lilga, Michael A.; Hallen, Richard T.; Albrecht, Karl O.; Cooper, Alan R.; Frye, John G.; Ramasamy, Karthikeyan Kallupalayam

    2018-04-03

    Systems, processes, and catalysts are disclosed for obtaining fuel and fuel blends containing selected ratios of open-chain and closed-chain fuel-range hydrocarbons suitable for production of alternate fuels including gasolines, jet fuels, and diesel fuels. Fuel-range hydrocarbons may be derived from ethylene-containing feedstocks and ethanol-containing feedstocks.

  18. Systems and processes for conversion of ethylene feedstocks to hydrocarbon fuels

    Lilga, Michael A.; Hallen, Richard T.; Albrecht, Karl O.; Cooper, Alan R.; Frye, John G.; Ramasamy, Karthikeyan Kallupalayam

    2017-09-26

    Systems, processes, and catalysts are disclosed for obtaining fuels and fuel blends containing selected ratios of open-chain and closed-chain fuel-range hydrocarbons suitable for production of alternate fuels including gasolines, jet fuels, and diesel fuels. Fuel-range hydrocarbons may be derived from ethylene-containing feedstocks and ethanol-containing feedstocks.

  19. Sol-gel process for thermal reactor fuel fabrication

    Mukerjee, S.K.

    2008-01-01

    Full text: Sol-gel processes have revolutionized conventional ceramic technology by providing extremely fine and uniform powders for the fabrication of ceramics. The use of this technology for nuclear fuel fabrication has also been explored in many countries. Unlike the conventional sol-gel process, sol-gel process for nuclear fuels tries to eliminate the preparation of powders in view of the toxic nature of the powders particularly those of plutonium and 233 U. The elimination of powder handling thus makes this process more readily amenable for use in glove boxes or for remote handling. In this process, the first step is the preparation of microspheres of the fuel material from a solution which is then followed by vibro-compaction of these microspheres of different sizes to obtain the required smear density of fuel inside a pin. The maximum achievable packing density of 92 % makes it suitable for fast reactors only. With a view to extend the applicability of sol-gel process for thermal reactor fuel fabrication the concept of converting the gel microspheres derived from sol-gel process, to the pellets, has been under investigation for several years. The unique feature of this process is that it combines the advantages of sol-gel process for the preparation of fuel oxide gel microspheres of reproducible quality with proven irradiation behavior of the pellet fuel. One of the important pre-requisite for the success of this process is the preparation of soft oxide gel microspheres suitable for conversion to dense pellets free from berry structure. Studies on the internal gelation process, one of the many variants of sol-gel process, for obtaining soft oxide gel microspheres suitable for gel pelletisation is now under investigation at BARC. Some of the recent findings related to Sol-Gel Microsphere Pelletisation (SGMP) in urania-plutonia and thoria-urania systems will be presented

  20. Thermal properties and phase transition in the fluoride, (NH4)3SnF7

    Kartashev, A.V.; Gorev, M.V.; Bogdanov, E.V.; Flerov, I.N.; Laptash, N.M.

    2016-01-01

    Calorimetric, dilatometric and differential thermal analysis studies were performed on (NH 4 ) 3 SnF 7 for a wide range of temperatures and pressures. Large entropy (δS 0 =22 J/mol K) and elastic deformation (δ(ΔV/V) 0 =0.89%) jumps have proven that the Pa-3↔Pm-3m phase transition is a strong first order structural transformation. A total entropy change of ΔS 0 =32.5 J/mol K is characteristic for the order–disorder phase transition, and is equal to the sum of entropy changes in the related material, (NH 4 ) 3 TiF 7 , undergoing transformation between the two cubic phases through the intermediate phases. Hydrostatic pressure decreases the stability of the high temperature Pm-3m phase in (NH 4 ) 3 SnF 7 , contrary to (NH 4 ) 3 TiF 7 , characterised by a negative baric coefficient. The effect of experimental conditions on the chemical stability of (NH 4 ) 3 SnF 7 was observed. - Graphical abstract: Strong first order structural transformation Pa-3↔Pm-3m in (NH 4 ) 3 SnF 7 is associated with very large total entropy change of ΔS 0 =32.5 J/mol K characteristic for the ordering processes and equal to the sum of entropy changes in the related (NH 4 ) 3 TiF 7 undergoing transformation between the same two cubic phases through the intermediate phases. - Highlights: • (NH 4 ) 3 SnF 7 undergoes strong first order Pa-3↔Pm-3m phase transition. • Anomalous behaviour of ΔC p and ΔV/V exists far below phase transition temperature. • Structural distortions are accompanied by huge total entropy change ΔS≈Rln50. • High pressure strongly increases the stability of Pa-3 phase in (NH 4 ) 3 SnF 7 . • Entropy of the Pa-3↔Pm-3m phase transition does not depend on pressure.

  1. Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

    Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

    2016-08-01

    The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel

  2. Spent nuclear fuels project characterization data quality objectives strategy

    Lawrence, L.A.; Thornton, T.A.; Redus, K.S.

    1994-12-01

    A strategy is presented for implementation of the Data Quality Objectives (DQO) process to the Spent Nuclear Fuels Project (SNFP) characterization activities. Westinghouse Hanford Company (WHC) and the Pacific Northwest Laboratory (PNL) are teaming in the characterization of the SNF on the Hanford Site and are committed to the DQO process outlined in this strategy. The SNFP characterization activities will collect and evaluate the required data to support project initiatives and decisions related to interim safe storage and the path forward for disposal. The DQO process is the basis for the activity specific SNF characterization requirements, termed the SNF Characterization DQO for that specific activity, which will be issued by the WHC or PNL organization responsible for the specific activity. The Characterization Plan prepared by PNL defines safety, remediation, and disposal issues. The ongoing Defense Nuclear Facility Safety Board (DNFSB) requirement and plans and the fuel storage and disposition options studies provide the need and direction for the activity specific DQO process. The hierarchy of characterization and DQO related documentation requirements is presented in this strategy. The management of the DQO process and the means of documenting the DQO process are described as well as the tailoring of the DQO process to the specific need of the SNFP characterization activities. This strategy will assure stakeholder and project management that the proper data was collected and evaluated to support programmatic decisions

  3. Galvanic cell for processing of used nuclear fuel

    Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

    2017-02-07

    A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

  4. Electrochemical fluorination for processing of used nuclear fuel

    Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

    2016-07-05

    A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

  5. Bugs and coal: processing fuels with biotechnology

    Shepard, M

    1987-06-01

    Bioprocessing of coal is developing along several fronts, each of potential significance to utilities. Researchers have found a fungus, polyporous versicolor, which can liquefy certain kinds of coal and scientists have genetically engineered bacteria that remove sulfur and ash-forming metal impurities from coal. Research programs are being undertaken to find organisms that will convert lignite into gaseous methane to produce gaseous fuel more economically than the current coal gasification methods. Researchers looking for ways to remove sulfur from coal before it is burned are evaluating the use of a bacterium called thiobacillus ferroxidans to enhance the physical removal of pyrite. 2 refs.

  6. Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes

    Dobrevski, I.; Zaharieva, N.

    2008-01-01

    Generally, Pressurized Water Reactor (WWER, PWR) Fuel Element Performance is connected with fuel cladding corrosion and crud deposition processes. By transient to extended fuel cycles in nuclear power reactors, aiming to achieve higher burnup and better fuel utilization, the role of these processes increases significantly. This evolution modifies the chemical and electrochemical conditions in the reactor primary system, including change of fuel claddings' environment. The higher duty cores are always attended with increased boiling (sub-cooled nucleate boiling) mainly on the feed fuel assemblies. This boiling process on fuel cladding surfaces can cause different consequences on fuel element cladding's environment characteristics. In the case of boiling at the cladding surfaces without or with some cover of corrosion product deposition, the behavior of gases dissolved in water phase is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. By these circumstances the concentrations of dissolved gases in cladding wall water layer can dramatically decrease, including also the case by which all dissolved gases to be stripped out. On the other hand it is known that the hydrogen is added to primary coolant in order to avoid the production of oxidants by radiolysis of water. It is clear that if boiling strips out dissolved hydrogen, the creation of oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen (O 2 ) and hydrogen (H 2 ) but also hydrogen peroxide (H 2 O 2 ) will be produced. While these hydrogen and oxygen will be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in wall water phase and will act as the most important factor for creation of oxidizing conditions in fuel

  7. Advanced fuel cycle on the basis of pyroelectrochemical process for irradiated fuel reprocessing and vibropacking technology

    Mayorshin, A.A.; Skiba, O.V.; Tsykanov, V.A.; Golovanov, V.N.; Bychkov, A.V.; Kisly, V.A.; Bobrov, D.A.

    2000-01-01

    For advanced nuclear fuel cycle in SSC RIAR there is developed the pyroelectrochemical process to reprocess irradiated fuel and produce granulated oxide fuel UO 2 , PuO 2 or (U,Pu)O 2 from chloride melts. The basic technological stage is the extraction of oxides as a crystal product with the methods either of the electrolysis (UO 2 and UO 2 -PuO 2 ) or of the precipitating crystalIization (PuO 2 ). After treating the granulated fuel is ready for direct use to manufacture vibropacking fuel pins. Electrochemical model for (U,Pu)O 2 coprecipitation is described. There are new processes being developed: electroprecipitation of mixed oxides - (U,Np)O 2 , (U,Pu,Np)O 2 , (U,Am)O 2 and (U,Pu,Am)O 2 . Pyroelectrochemical production of mixed actinide oxides is used both for reprocessing spent fuel and for producing actinide fuel. Both the efficiency of pyroelectrochemical methods application for reprocessing nuclear fuel and of vibropac technology for plutonium recovery are estimated. (author)

  8. Research on plant of metal fuel fabrication using casting process

    Senda, Yasuhide; Mori, Yukihide

    2003-12-01

    This document presents the plant concept of metal fuel fabrication system (38tHM/y) using casting process in electrolytic recycle, which based on recent studies of its equipment design and quality control system. And we estimate the cost of its construction and operation, including costs of maintenance, consumed hardware and management of waste. The content of this work is as follows. (1) Designing of fuel fabrication equipment: We make material flow diagrams of the fuel fabrication plant and rough designs of the injection casting furnace, demolder and inspection equipment. (2) Designing of resolution system of liquid waste, which comes from analytical process facility. Increased analytical items, we rearrange analytical process facility, estimate its chemicals and amount of waste. (3) Arrangement of equipments: We made a arrangement diagram of the metal fuel fabrication equipments in cells. (4) Estimation of cost data: We estimated cost to construct the facility and to operate it. (author)

  9. Production process and quality control for the HTTR fuel

    Yoshimuta, S.; Suzuki, N.; Kaneko, M.; Fukuda, K.

    1991-01-01

    Development of the production and inspection technology for High Temperature Engineering Test Reactor (HTTR) fuel has been carried out by cooperative work between Japan Atomic Energy Research Institute (JAERI) and Nuclear Fuel Industries, Ltd (NFI). The performance and the quality level of the developed fuel are well established to meet the design requirements of the HTTR. For the commercial scale production of the fuel, statistical quality control and quality assurance must be carefully considered in order to assure the safety of the HTTR. It is also important to produce the fuel under well controlled process condition. To meet these requirements in the production of the HTTR fuel, a new production process and quality control system is to be introduced in the new facilities. The main feature of the system is a computer integrated control system. Process control data at each production stage of products and semi-products are all gathered by terminal computers and processed by a host computer. The processed information is effectively used for the production, quality and accountancy control. With the aid of this system, all the products will be easily traceable from starting materials to final stages and the statistical evaluation of the quality of products becomes more reliable. (author). 8 figs

  10. IFR fuel cycle process equipment design environment and objectives

    Rigg, R.H.

    1993-01-01

    Argonne National laboratory (ANL) is refurbishing the hot cell facility originally constructed with the EBR-II reactor. When refurbishment is complete, the facility win demonstrate the complete fuel cycle for current generation high burnup metallic fuel elements. These are sodium bonded, stainless steel clad fuel pins of U-Zr or U-Pu-Zr composition typical of the fuel type proposed for a future Integral Fast Reactor (IFR) design. To the extent possible, the process equipment is being built at full commercial scale, and the facility is being modified to incorporate current DOE facility design requirements and modem remote maintenance principles. The current regulatory and safety environment has affected the design of the fuel fabrication equipment, most of which will be described in greater detail in subsequent papers in this session

  11. Model studying the processes arising during fuel element overheating

    Usynin, G.B.; Anoshkin, Yu.I.; Vlasichev, G.N.; Galitskikh, Yu.N.; Semenychev, M.A.

    1986-01-01

    A calculational technique for studying heating and melting of fuel elements in the BN type reactors during an accident with heat release failure and a simulator with central rod heater intended for out-of-pile experiments is developed. The time rangeof the characteristic melting steps for the most thermally stressed fuel element at the reactor nominal power is calculated. The experimental study of the fuel element melting using a simulator with a tungsten heater has proved that the technique for the simultor and fuel can melting, respectively, is correct. The developed technique is used for determining the geometrical values and operational conditions for experiments with simulators, when quantitative and qualitative characteristics of the process under study are rather close to those natural for fuel elements

  12. Nuclear reactor fuel cycle technology with pyroelectrochemical processes

    Skiba, O.V.; Maershin, A.A.; Bychkov, A.V.; Zhdanov, A.N.; Kislyj, V.A.; Vavilov, S.K.; Babikov, L.G.

    1999-01-01

    A group of dry technologies and processes of vibro-packing granulated fuel in combination with unique properties of vibro-packed FEs make it possible to implement a new comprehensive approach to the fuel cycle with plutonium fuel. Testing of a big number of FEs with vibro-packed U-Pu oxide fuel in the BOR-60 reactor, successful testing of experimental FSAs in the BN-600 rector, reliable operation of the experimental and research complex facilities allow to make the conclusion about a real possibility to develop a safe, economically beneficial U-Pu fuel cycle based on the technologies enumerated above and to use both reactor-grade and weapon-grade plutonium in nuclear reactors with a reliable control and accounting system [ru

  13. Development of Voloxidation Process for Treatment of LWR Spent Fuel

    Park, J. J.; Jung, I. H.; Shin, J. M. (and others)

    2007-08-15

    The objective of the project is to develop a process which provides a means to recover fuel from the cladding, and to simplify downstream processes by recovering volatile fission products. This work focuses on the process development in three areas ; the measurement and assessment of the release behavior for the volatile and semi-volatile fission products from the voloxidation process, the assessment of techniques to trap and recover gaseous fission products, and the development of process cycles to optimize fuel cladding separation and fuel particle size. High temperature adsorption method of KAERI was adopted in the co-design of OTS for hot experiment in INL. KAERI supplied 6 sets of filter for hot experiment. Three hot experiment in INL hot cell from the 25th of November for two weeks with attaching 4 KAERI staffs had been carried out. The results were promising. For example, trapping efficiency of Cs was 95% and that of I was 99%, etc.

  14. Analysis of the ATR fuel element swaging process

    Richins, W.D.; Miller, G.K.

    1995-12-01

    This report documents a detailed evaluation of the swaging process used to connect fuel plates to side plates in Advanced Test Reactor (ATR) fuel elements. The swaging is a mechanical process that begins with fitting a fuel plate into grooves in the side plates. Once a fuel plate is positioned, a lip on each of two side plate grooves is pressed into the fuel plate using swaging wheels to form the joints. Each connection must have a specified strength (measured in terms, of a pullout force capacity) to assure that these joints do not fail during reactor operation. The purpose of this study is to analyze the swaging process and associated procedural controls, and to provide recommendations to assure that the manufacturing process produces swaged connections that meet the minimum strength requirement. The current fuel element manufacturer, Babcock and Wilcox (B ampersand W) of Lynchburg, Virginia, follows established procedures that include quality inspections and process controls in swaging these connections. The procedures have been approved by Lockheed Martin Idaho Technologies and are designed to assure repeatability of the process and structural integrity of each joint. Prior to July 1994, ATR fuel elements were placed in the Hydraulic Test Facility (HTF) at the Idaho National Engineering Laboratory (AGNAIL), Test Reactor Area (TRA) for application of Boehmite (an aluminum oxide) film and for checking structural integrity before placement of the elements into the ATR. The results presented in this report demonstrate that the pullout strength of the swaged connections is assured by the current manufacturing process (with several recommended enhancements) without the need for- testing each element in the HTF

  15. Probabilistic Risk Assessment on Maritime Spent Nuclear Fuel Transportation

    Christian, Robby; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    Spent nuclear fuel (SNF) management has been an indispensable issue in South Korea. Before a long term SNF solution is implemented, there exists the need to distribute the spent fuel pool storage loads. Transportation of SNF assemblies from populated pools to vacant ones may preferably be done through the maritime mode since all nuclear power plants in South Korea are located at coastal sites. To determine its feasibility, it is necessary to assess risks of the maritime SNF transportation. This work proposes a methodology to assess the risk arising from ship collisions during the transportation of SNF by sea. Its scope is limited to the damage probability of SNF packages given a collision event. The effect of transport parameters' variation to the package damage probability was investigated to obtain insights into possible ways to minimize risks. A reference vessel and transport cask are given in a case study to illustrate the methodology's application.

  16. Storage of spent nuclear fuel: the problem of spent nuclear fuel in Bulgaria

    Boyadzhiev, Z.; Vapirev, E.

    1995-01-01

    A review of existing technologies for wet and dry storage of spent nuclear fuel (SNF) and the reprocessing policies is presented. The problem of SNF in Bulgaria is arising from nonobservance of the obligation to return SNF back to the former Soviet Union as agreed in the construction contract. In November 1994 approximately 1800 fuel assemblies have been stored in away-from-reactor (AFR) facility and another 1060 in at-reactor (AR) pools. The national policy is to export SNF out of the country. The AFR facility has a limited capacity and it is designed only for WWER-440 fuel although work is going on to extend it in order to store WWER-1000 SNF. 14 refs

  17. Storage of spent nuclear fuel: the problem of spent nuclear fuel in Bulgaria

    Boyadzhiev, Z; Vapirev, E [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1996-12-31

    A review of existing technologies for wet and dry storage of spent nuclear fuel (SNF) and the reprocessing policies is presented. The problem of SNF in Bulgaria is arising from nonobservance of the obligation to return SNF back to the former Soviet Union as agreed in the construction contract. In November 1994 approximately 1800 fuel assemblies have been stored in away-from-reactor (AFR) facility and another 1060 in at-reactor (AR) pools. The national policy is to export SNF out of the country. The AFR facility has a limited capacity and it is designed only for WWER-440 fuel although work is going on to extend it in order to store WWER-1000 SNF. 14 refs.

  18. Process for the fabrication of a nuclear fuel

    Hirose, Yasuo.

    1970-01-01

    Herein disclosed is a process for fabricating a nuclear fuel incorporating either uranium or plutonium. A pellet-like substrate consisting of a packed powder ceramic fuel such as uranium or plutonium is prepared with the horizontal surface of the body provided with a masking. Next, after impregnating the substrate voids with a solution consisting of a fissile material or mixture of fissile material and poison, the solvent is removed by a chemical deposition process which causes the impregnated material to migrate through capillary action toward the vicinity of the fuel body surface. Sintering and pyrolysis of the deposited material and masking are subsequently carried out to yield a fuel body having adjacent to its surface an intensely concentrated layer of either fissile material or a mixture of fissile material and poison. (Owens, K.J.)

  19. On board fuel processing for using in electric vehicles

    Paez, Daniel E.; Marquez, Marco A.

    1999-01-01

    The increase in vehicle population, the emission effects upon the environment, and the growing concern of industrialized nations to reduce oil dependency, are the arguments for the new developments that may change the automobile revolution within the next decades. However, the electricity to move the future vehicles must come from the processing of liquid fuels on board. Liquid fuels such as gasoline have the advantage of having good on-site system for distribution and supply directly to the vehicle and will compete for staying as the energy source of the future. What are the opportunities in R and D and how to take advantage of them are analyzed in this document. Liquid fuel processing technologies and fuel options are also described by PDVSA-INTEVEP

  20. DESIGN OF A CONCRETE SLAB FOR STORAGE OF SNF AND HLW CASKS

    J. Bisset

    2005-01-01

    This calculation documents the design of the Spent Nuclear Fuel (SNF) and High-Level Waste (HLW) Cask storage slab for the Aging Area. The design is based on the weights of casks that may be stored on the slab, the weights of vehicles that may be used to move the casks, and the layout shown on the sketch for a 1000 Metric Ton of Heavy Metal (MTHM) storage pad on Attachment 2, Sht.1 of the calculation 170-C0C-C000-00100-000-00A (BSC 2004a). The analytical model used herein is based on the storage area for 8 vertical casks. To simplify the model, the storage area of the horizontal concrete modules and their related shield walls is not included. The heavy weights of the vertical storage casks and the tensile forces due to pullout at the anchorages will produce design moments and shear forces that will envelope those that would occur in the storage area of the horizontal modules. The design loadings will also include snow and live loads. In addition, the design will also reflect pertinent geotechnical data. This calculation will document the preliminary thickness and general reinforcing steel requirements for the slab. This calculation also documents the initial design of the cask anchorage. Other slab details are not developed in this calculation. They will be developed during the final design process. The calculation also does not include the evaluation of the effects of cask drop loads. These will be evaluated in this or another calculation when the exact cask geometry is known

  1. Capabilities for processing shipping casks at spent fuel storage facilities

    Baker, W.H.; Arnett, L.M.

    1978-01-01

    Spent fuel is received at a storage facility in heavily shielded casks transported either by rail or truck. The casks are inspected, cooled, emptied, decontaminated, and reshipped. The spent fuel is transferred to storage. The number of locations or space inside the building provided to perform each function in cask processing will determine the rate at which the facility can process shipping casks and transfer spent fuel to storage. Because of the high cost of construction of licensed spent fuel handling and storage facilities and the difficulty in retrofitting, it is desirable to correctly specify the space required. In this paper, the size of the cask handling facilities is specified as a function of rate at which spent fuel is received for storage. The minimum number of handling locations to achieve a given throughput of shipping casks has been determined by computer simulation of the process. The simulation program uses a Monte Carlo technique in which a large number of casks are received at a facility with a fixed number of handling locations in each process area. As a cask enters a handling location, the time to process the cask at that location is selected at random from the distribution of process time. Shipping cask handling times are based on experience at the General Electric Storage Facility, Morris, Illinois. Shipping cask capacity is based on the most recent survey available of the expected capability of reactors to handle existing rail or truck casks

  2. Process and equipment for locating defective fuel rods of a reactor fuel element

    Jester, A.; Honig, H.

    1977-01-01

    By this equipment, well-known processes for determining defective fuel rods of a reactor fuel element are improved in such a fashion that defective fuel rods can be located individually, so that it is possible to replace them. The equipment consists of a cylindrical test vessel open above, which accommodates the element to be tested, so that an annular space is left between the latter's external circumference and the wall of the vessel, and so that the fuel rods project above the vessel. A bell in the shape of a frustrum of a cone is inverted over the test vessel, which has an infra-red measuring equipment at a certain distance above the tops of the fuel rods. The fuel element to be tested together with the test vessel and hood are immersed in a basin full of water, which displaces water by means of gas from the hood. The post-shutdown heat increases the temperature in the water space of the test vessel, which is stabilised at 100 0 C. In each defective fuel rod the water which has penetrated the defective fuel rod previously, or does so now, starts to boil. The steam rising in the fuel rod raises the temperature of the defective fuel rod compared to all the sound ones. The subsequent measurement easily determines this. Where one can expect interference with the measurement by appreciable amounts of gamma rays, the measuring equipment is removed from the path of radiation by mirror deflection in a suitably shaped measuring hood. (FW) [de

  3. Parallel processing of neutron transport in fuel assembly calculation

    Song, Jae Seung

    1992-02-01

    Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's

  4. Alteration to the SWI/SNF complex in human cancers

    Vanessa S. Gordon

    2011-12-01

    Full Text Available The SWI/SNF complex is a key catalyst for gene expression and regulates a variety of pathways, many of which have anticancer roles. Its central roles in cellular growth control, DNA repair, differentiation, cell adhesion and development are often targeted, and inactivated, during cancer development and progression. In this review, we will discuss what is known about how SWI/SNF is inactivated, and describe the potential impact of abrogating this complex. BRG1 and BRM are the catalytic subunits which are essential for SWI/SNF function, and thus, it is not surprising that they are lost in a variety of cancer types. As neither gene is mutated when lost, the mechanism of suppression, as well as the impact of potential gene activity restoration, are reviewed.

  5. New glass material oxidation and dissolution system facility: Direct conversion of surplus fissile materials, spent nuclear fuel, and other material to high-level-waste glass. Storage and disposition of weapons-usable fissile materials programmatic environmental impact statement data report: Predecisional draft

    Forsberg, C.W.; Elam, K.R.; Reich, W.J.

    1995-01-01

    With the end of the Cold War, countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. It has been recommended that these surplus fissile materials (SFMs) be processed so that they are no more accessible than plutonium in spent nuclear fuel (SNF). This SNF standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. This report provides for the PEIS the necessary input data on a new method for the disposition of SFMs: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptunium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal. The primary form of this SNF is Hanford-N SNF with preirradiation uranium enrichments between 0.95 and 1.08%. The final product is a plutonium, low-enriched-uranium, HLW, borosilicate glass for disposition in a geological repository. The proposed conversion process is the Glass Material Oxidation and Dissolution System (GMODS), which is a new process. The initial analysis of the GMODS process indicates that a MODS facility for this application would be similar in size and environmental impact to the Defense Waste Processing Facility (DWPF) at the Savannah River Site. Because of this, the detailed information available on DWPF was used as the basis for much of the GMODS input into the SFMs PEIS

  6. Demonstration of pyrometallurgical processing for metal fuel and HLW

    Tadafumi, Koyama; Kensuke, Kinoshita; Takatoshi, Hizikata; Tadashi, Inoue; Ougier, M.; Rikard, Malmbeck; Glatz, J.P.; Lothar, Koch

    2001-01-01

    CRIEPI and JRC-ITU have started a joint study on pyrometallurgical processing to demonstrate the capability of this type of process for separating actinide elements from spent fuel and HLW. The equipment dedicated for this experiments has been developed and installed in JRC-ITU. The stainless steel box equipped with tele-manipulators is operated under pure Ar atmosphere, and prepared for later installation in a hot cell. Experiments on pyro-processing of un-irradiated U-Pu-Zr metal alloy fuel by molten salt electrorefining has been carried out. Recovery of U and Pu from this type alloy fuel was first demonstrated with using solid iron cathode and liquid Cd cathode, respectively. (author)

  7. Analysis of the processes defining radionuclide migration from deep geological repositories in porous medium

    Brazauskaite, A.; Poskas, P.

    2004-01-01

    Due to the danger of exposure arising from long-lived radionuclides to humans and environment, spent nuclear fuel (SNF) and high level waste (HLW) are not allowed to be disposed of in near surface repositories. There exists an international consensus that such high level and long-lived radioactive wastes are best disposed of in geological repositories using a system of engineered and natural barriers. At present, the geological repository of SNF and HLW has not been realized yet in any country but there is a lot of experience in the assessment of radionuclide migration from deep repositories, investigations of different processes related to the safety of a disposal system. The aim of this study was to analyze the processes related to the radionuclide migration from deep geological repositories in porous medium such as SNF matrix dissolution, release mechanism of radionuclides from SNF matrix, radionuclide solubility, sorption, diffusive, advective transport of radionuclides from the canister and through the engineered and natural barriers. It has been indicated that SNF matrix dissolution, radionuclide solubility and sorption are sensitive to ambient conditions prevailing in the repository. The approaches that could be used for modeling the radionuclide migration from deep repositories in porous medium are also presented. (author)

  8. Processing of carbon composite paper as electrode for fuel cell

    Mathur, R.B.; Maheshwari, Priyanka H.; Dhami, T.L. [Carbon Technology Unit, National Physical Laboratory, New Delhi 110012 (India); Sharma, R.K.; Sharma, C.P. [Soft Polymeric Group, Division of Engineering Materials, National Physical Laboratory, New Delhi 110012 (India)

    2006-10-27

    The porous carbon electrode in a fuel cell not only acts as an electrolyte and a catalyst support, but also allows the diffusion of hydrogen fuel through its fine porosity and serves as a current-carrying conductor. A suitable carbon paper electrode is developed and possesses the characteristics of high porosity, permeability and strength along with low electrical resistivity so that it can be effectively used in proton-exchange membrane and phosphoric acid fuel cells. The electrode is prepared through a combination of two important techniques, viz., paper-making technology by first forming a porous chopped carbon fibre preform, and composite technology using a thermosetting resin matrix. The study reveals an interdependence of one parameter on another and how judicious choice of the processing conditions are necessary to achieve the desired characteristics. The current-voltage performance of the electrode in a unit fuel cell matches that of a commercially-available material. (author)

  9. Quality control in the fuel elements production process

    Katanic-Popovic, J.; Spasic, Z.; Djuricis, Lj.

    1977-01-01

    Recently great attention has been paid at the international level to the analysis of production processes and quality control of fuel and fuel elements with the aim to speed up activity of proposing and accepting standards and measurement methods. IAEA also devoted great interest to these problems appealing to more active participation of all users and producers fuel elements in a general effort to secure successful work of nuclear plants. For adequate and timely participation in future in the establishment and analysis of general requirements and documentation for the control of purchased or self produced fuel elements in out country it is necessary to be well informed and to follow this activity at the international level. (author)

  10. Process and apparatus for sealing nuclear reactor fuel

    Duncan, R.; Barna, R.P.

    1978-01-01

    A process and apparatus for simultaneously pressurizing a fuel rod having a plug in one end, welding a plug in the other end and sealing a gas pressurizing orifice therein in a single operation is described. A weld chamber is provided which accommodates one end of a seal rod having a plug fixed in the rod end by a friction fit. A mechanism pushes the fuel rod into the weld chamber which is then pressurized to force gas through a plug orifice into the fuel rod. During subsequent rotation of the rod, an electrode in the weld chamber forms a weld puddle which bridges the end plug-fuel rod interface and the plug orifice to thereby weld the plug in the rod and seal the plug orifice in a single operation. 6 claims, 3 figures

  11. Advanced Fuels and Combustion Processes for Propulsion

    2010-09-01

    production from biomass steam reforming – Conduct a feasibility analysis of the proposed integrated process Energia Technologies - D. Nguyen & K. Parimi...strength foam material development by Ultramet – Combustion experiments performed U. Of Alabama – End-user input provided by Solar Turbines Major

  12. PHWR fuel fabrication with imported uranium - procedures and processes

    Rao, R.V.R.L.V.; Rameswara Rao, A.; Hemantha Rao, G.V.S.; Jayaraj, R.N.

    2010-01-01

    Following the 123 agreement and subsequent agreements with IAEA & NSG, Government of India has entered into bilateral agreements with different countries for nuclear trade. Department of Atomic Energy (DAE), Government of India, has entered into contract with few countries for supply of uranium material for use in the safeguarded PHWRs. Nuclear Fuel Complex (NFC), an industrial unit of DAE, established in the early seventies, is engaged in the production of Nuclear Fuel and Zircaloy items required for Nuclear Power Reactors operating in the country. NFC has placed one of its fuel fabrication facilities (NFC, Block-A, INE-) under safeguards. DAE has opted to procure uranium material in the form of ore concentrate and fuel pellets. Uranium ore concentrate was procured as per the ASTM specifications. Since no international standards are available for PHWR fuel pellets, Specifications have to be finalized based on the present fabrication and operating experience. The process steps have to be modified and fine tuned for handling the imported uranium material especially for ore concentrate. Different transportation methods are to be employed for transportation of uranium material to the facility. Cost of the uranium material imported and the recoveries at various stages of fuel fabrication have impact on the fuel pricing and in turn the unit energy costs. Similarly the operating procedures have to be modified for safeguards inspections by IAEA. NFC has successfully manufactured and supplied fuel bundles for the three 220 MWe safeguarded PHWRs. The paper describes various issues encountered while manufacturing fuel bundles with different types of nuclear material. (author)

  13. Digital mock-up for the spent fuel disassembly processes

    Lee, J. Y.; Kim, S. H.; Song, T. G.; Kim, Y. H.; Hong, D. H.; Yoon, J. S.

    2000-12-01

    In this study, the graphical design system is developed and the digital mock-up is implemented for designing the spent fuel handling and disassembly processes. The system consists of a 3D graphical modeling system, a devices assembling system, and a motion simulation system. This system is used throughout the design stages from the conceptual design to the motion analysis. By using this system, all the process involved in the spent fuel handling and disassembly processes are analyzed and optimized. Also, this system is used in developing the on-line graphic simulator which synchronously simulates the motion of the equipment in a real time basis by connecting the device controllers with the graphic server through the TCP/IP network. This simulator can be effectively used for detecting the malfunctions of the process equipment which is remotely operated. Thus, the simulator enhances the reliability and safety of the spent fuel handling process by providing the remote monitoring function of the process. The graphical design system and the digital mock-up system can be effectively used for designing the process equipment, as well as the optimized process and maintenance process. And the on-line graphic simulator can be an alternative of the conventional process monitoring system which is a hardware based system

  14. Process for production of liquid fuels

    1947-03-14

    The process comprises continually passing a current of carbonaceous material, finely divided in a vertical zone of reaction, finely divided carbonaceous material raised to a fluidized state dispersed in gas flowing from the bottom to the top carrying into the zone of reaction an oxidizing gas in sufficient quantity to supply the heat for combustion by burning part of the carbonaceous material withdrawing from the zone of reaction the products of the combustible fluids, withdrawing from the zone the solid carbonaceous products of the reaction and reacting a part of the lesser products left in a second zone of reaction by means of steam to produce oxides of carbon and hydrogen.

  15. Standardization of Fat:SNF ratio of milk and addition of sprouted wheat fada (semolina) for the manufacture of halvasan.

    Chaudhary, Apurva H; Patel, H G; Prajapati, P S; Prajapati, J P

    2015-04-01

    Traditional Indian Dairy Products such as Halvasan are manufactured in India using an age old practice. For manufacture of such products industrially, a standard formulation is required. Halvasan is a region specific, very popular heat desiccated milk product but has not been studied scientifically. Fat and Solids-not-fat (SNF) plays an important role in physico-chemical, sensory, textural characteristics and also the shelf life of any milk sweet. Hence for process standardization of Halvasan manufacture, different levels of Fat:SNF ratios i.e. 0.44, 0.55, 0.66 and 0.77 of milk were studied so that an optimum level yielding best organoleptic characteristics in final product can be selected. The product was made from milk standardized to these ratios of Fat:SNF and the product was manufactured as per the method tentatively employed on the basis of characterization of market samples of the product in laboratory. Based on the sensory results obtained, a Fat:SNF ratio of 0.66 for the milk has been selected. In the similar way, for standardizing the rate of addition of fada (semolina); 30, 40, 50 and 60 g fada (semolina) per kg of milk were added and based on the sensory observations, the level of fada (semolina) addition @50 gm/kg of milk was adjudged the best for Halvasan manufacture and hence selected.

  16. Fuel sparing: Control of industrial furnaces using process gas as supplemental fuel

    Boisvert, Patrick G.; Runstedtler, Allan

    2014-01-01

    Combustible gases from industrial processes can be used to spare purchased fuels such as natural gas and avoid wasteful flaring of the process gases. One of the challenges of incorporating these gases into other furnaces is their intermittent availability. In order to incorporate the gases into a continuously operating furnace, the furnace control system must be carefully designed so that the payload is not affected by the changing fuel. This paper presents a transient computational fluid dynamics (CFD) model of an industrial furnace that supplements natural gas with carbon monoxide during furnace operation. A realistic control system of the furnace is simulated as part of the CFD calculation. The time dependent changes in fuels and air injection on the furnace operation is observed. It is found that there is a trade-off between over-controlling the furnace, which results in too sensitive a response to normal flow oscillations, and under-controlling, which results in a lagged response to the fuel change. - Highlights: •Intermittently available process gases used in a continuously operating furnace. •Study shows a trade-off between over-controlling and under-controlling the furnace. •Over-controlling: response too sensitive to normal flow oscillations. •Under-controlling: lagged response to changing fuel composition. •Normal flow oscillations in furnace would not be apparent in steady-state model

  17. K-Basin spent nuclear fuel characterization data report

    Abrefah, J.; Gray, W.J.; Ketner, G.L.; Marschman, S.C.; Pyecha, T.D.; Thornton, T.A.

    1995-11-01

    The spent nuclear fuel (SNF) project characterization activities will be furnishing technical data on SNF stored at the K Basins in support of a pathway for placement of a ''stabilized'' form of SNF into an interim storage facility. This report summarizes the results so far of visual inspection of the fuel samples, physical characterization (e.g., weight and immersion density measurements), metallographic examinations, and controlled atmosphere furnace testing of three fuel samples shipped from the KW Basin to the Postirradiation Testing Laboratory (PTL). Data on sludge material collected by filtering the single fuel element canister (SFEC) water are also discussed in this report

  18. Spent nuclear fuel project product specification

    Pajunen, A.L.

    1998-01-01

    Product specifications are limits and controls established for each significant parameter that potentially affects safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for transport to dry storage. The product specifications in this document cover the spent fuel packaged in MultiCanister Overpacks (MCOs) to be transported throughout the SNF Project. The SNF includes N Reactor fuel and single-pass reactor fuel. The FRS removes the SNF from the storage canisters, cleans it, and places it into baskets. The MCO loading system places the baskets into MCO/Cask assembly packages. These packages are then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the MCO cask packages are transferred to the Canister Storage Building (CSB), where the MCOs are removed from the casks, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The key criteria necessary to achieve these goals are documented in this specification

  19. Process for producing a fuel suitable for degassing from refuse

    Sulzberger, J

    1975-11-20

    Utilization of the heat energy of refuse in waste incineration plants is time-consuming and expensive due to high investment and operation costs. The inventor recommends to process the refuse to a sterile, handy and storable fuel. For this propose the refuse should be crushed, kneaded and pressed. The briquettes produced in this way should be dried.

  20. 42 CFR 424.20 - Requirements for posthospital SNF care.

    2010-10-01

    ... statements may be signed by— (1) The physician responsible for the case or, with his or her authorization, by a physician on the SNF staff or a physician who is available in case of an emergency and has knowledge of the case; or (2) A nurse practitioner or clinical nurse specialist, neither of whom has a...

  1. Progress in realization of the state policy in RW and SNF Management in the Russian Federation

    Borzunov, Andrey I.

    1999-01-01

    The basic infrastructure at the majority of the enterprises for management of radioactive waste (RW) and spent nuclear fuel (SNF) built in Russia in the 1960s and 1970s are now morally and technically obsolete and require reconstruction. As stated in this presentation, the most complicated problem is the shortage of financial resources, and International support is very important. The presentation is organised in sections discussing (1) the problem, (2) basic aspects of the State policy in this field, (3) the federal institutions in charge, (4) the principles upon which the State policy is grounded, (5) the main objectives of the RW and SNF management in Russia, (6) the federal programme: Radioactive wastes and spent nuclear materials management, their disposal and burial for the period 1996-2005, (7) plans for impending solution of the problems of the Northern and Pacific regions of Russia, (8) some top priority work of Minatom, (9) measures planned at the Russian power plants, (10) some basic results so far, (11) international co-operation

  2. Progress in realization of the state policy in RW and SNF Management in the Russian Federation

    Borzunov, Andrey I

    1999-07-01

    The basic infrastructure at the majority of the enterprises for management of radioactive waste (RW) and spent nuclear fuel (SNF) built in Russia in the 1960s and 1970s are now morally and technically obsolete and require reconstruction. As stated in this presentation, the most complicatedproblem is the shortage of financial resources, and International support is very important. The presentation is organised in sections discussing (1) the problem, (2) basic aspects of the State policy in this field, (3) the federal institutions in charge, (4) the principles upon which the State policy is grounded, (5) the main objectives of the RW and SNF management in Russia, (6) the federal programme: Radioactive wastes and spent nuclear materials management, their disposal and burial for the period 1996-2005, (7) plans for impending solution of the problems of the Northern and Pacific regions of Russia, (8) some top priority work of Minatom, (9) measures planned at the Russian power plants, (10) some basic results so far, (11) international co-operation.

  3. Handbook on process and chemistry of nuclear fuel reprocessing. 3rd edition

    2015-03-01

    The fundamental data on spent nuclear fuel reprocessing and related chemistry was collected and summarized as a new edition of 'Handbook on Process and Chemistry of Nuclear Fuel Reprocessing'. The purpose of this handbook is contribution to development of the fuel reprocessing and fuel cycle technology for uranium fuel and mixed oxide fuel utilization. Contents in this book was discussed and reviewed by specialists of science and technology on fuel reprocessing in Japan. (author)

  4. Status of US program for disposal of spent nuclear fuel

    Smith, R.I.

    1991-04-01

    In this paper, a brief history of the United States' program for the disposal of spent nuclear fuel (SNF) and the legislative acts that have guided the program are discussed. The current plans and schedules for beginning acceptance of SNF from the nuclear utilities for disposal are described, and some of the development activities supporting the program are discussed. And finally, the viability of the SNF disposal fee presently paid into the Nuclear Waste Fund by the owners/generators of commercial SNF and high-level waste (HLW) is examined. 12 refs., 9 figs

  5. Process for gasifying fuels with the recovery of rich gases

    Jahns, F

    1921-04-10

    A process for gasifying fuels with recovery of water-free, rich-in-tar gases in a ring-gas-producer characterized by hot-gas-stream arising from the gasification bed of a fresh chamber in the known way is divided. One part is conducted through an old chamber, the other part is led first during the drying through the fresh fuel and with the received water-vapor also through the old chamber and then during the carbonization with the carbonization products is led to the carbonization-gas conduit.

  6. Structural analysis of advanced spent fuel conditioning process

    Gu, J. H.; Jung, W. M.; Jo, I. J.; Gug, D. H.; Yoo, K. S.

    2003-01-01

    An advanced spent fuel conditioning process (ACP) is developing for the safe and effective management of spent fuels which arising from the domestic nuclear power plants. And its demonstration facility is under design. This facility will be prepared by modifying IMEF's reserve hot cell facility which reserved for future usage by considering the characteristics of ACP. This study presents a basic structural architecture design and analysis results of ACP hot cell including modification of the IMEF. The results of this study will be used for the detail design of ACP demonstration facility, and utilized as basic data for the licensing of the ACP facility

  7. Preliminary Hanford technical input for the Department of Energy programmatic spent nuclear fuel management and Idaho National Engineering Laboratory environmental restoration and waste management programs environmental impact statement

    Bergsman, K.H.

    1995-03-01

    The US Department of Energy (DOE) is currently evaluating its programmatic options for the safe management of its diverse spent nuclear fuel (SNF) inventory in the Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Environmental Impact Statement (SNF and INEL EIS). In the SNF and INEL EIS, the DOE is assessing five alternatives for SNF management, which consider at which of the DOE sites each of the various SNF types should be managed until ultimate disposition. The range of SNF inventories considered for management at the Hanford Site in the SNF and INEL EIS include the current Hanford Site inventory, only the current Hanford Site defense production SNF inventory, the DOE complex-wide SNF inventory, or none at all. Site-specific SNF management decisions will be evaluated in separate National Environmental Policy Act evaluations. Appendixes A and B include information on (1) additional facilities required to accommodate inventories of SNF within each management alternative, (2) existing and new SNF management facility descriptions, (3) facility costs for construction and operation, (4) facility workforce requirements for construction and operation, and (5) facility discharges. The information was extrapolated from existing analyses to the extent possible. New facility costs, manpower requirements, and similar information are based on rough-order-of-magnitude estimates

  8. Method For Processing Spent (Trn,Zr)N Fuel

    Miller, William E.; Richmann, Michael K.

    2004-07-27

    A new process for recycling spent nuclear fuels, in particular, mixed nitrides of transuranic elements and zirconium. The process consists of two electrorefiner cells in series configuration. A transuranic element such as plutonium is reduced at the cathode in the first cell, zirconium at the cathode in the second cell, and nitrogen-15 is released and captured for reuse to make transuranic and zirconium nitrides.

  9. Candida albicans Swi/Snf and Mediator Complexes Differentially Regulate Mrr1-Induced MDR1 Expression and Fluconazole Resistance.

    Liu, Zhongle; Myers, Lawrence C

    2017-11-01

    Long-term azole treatment of patients with chronic Candida albicans infections can lead to drug resistance. Gain-of-function (GOF) mutations in the transcription factor Mrr1 and the consequent transcriptional activation of MDR1 , a drug efflux coding gene, is a common pathway by which this human fungal pathogen acquires fluconazole resistance. This work elucidates the previously unknown downstream transcription mechanisms utilized by hyperactive Mrr1. We identified the Swi/Snf chromatin remodeling complex as a key coactivator for Mrr1, which is required to maintain basal and induced open chromatin, and Mrr1 occupancy, at the MDR1 promoter. Deletion of snf2 , the catalytic subunit of Swi/Snf, largely abrogates the increases in MDR1 expression and fluconazole MIC observed in MRR1 GOF mutant strains. Mediator positively and negatively regulates key Mrr1 target promoters. Deletion of the Mediator tail module med3 subunit reduces, but does not eliminate, the increased MDR1 expression and fluconazole MIC conferred by MRR1 GOF mutations. Eliminating the kinase activity of the Mediator Ssn3 subunit suppresses the decreased MDR1 expression and fluconazole MIC of the snf2 null mutation in MRR1 GOF strains. Ssn3 deletion also suppresses MDR1 promoter histone displacement defects in snf2 null mutants. The combination of this work with studies on other hyperactive zinc cluster transcription factors that confer azole resistance in fungal pathogens reveals a complex picture where the induction of drug efflux pump expression requires the coordination of multiple coactivators. The observed variations in transcription factor and target promoter dependence of this process may make the search for azole sensitivity-restoring small molecules more complicated. Copyright © 2017 American Society for Microbiology.

  10. Fuel production from coal by the Mobil Oil process using nuclear high-temperature process heat

    Hoffmann, G.

    1982-01-01

    Two processes for the production of liquid hydrocarbons are presented: Direct conversion of coal into fuel (coal hydrogenation) and indirect conversion of coal into fuel (syngas production, methanol synthesis, Mobil Oil process). Both processes have several variants in which nuclear process heat may be used; in most cases, the nuclear heat is introduced in the gas production stage. The following gas production processes are compared: LURGI coal gasification process; steam reformer methanation, with and without coal hydrogasification and steam gasification of coal. (orig./EF) [de

  11. Spent nuclear fuel project integrated schedule plan

    Squires, K.G.

    1995-01-01

    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

  12. Spent nuclear fuel project integrated schedule plan

    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.

  13. Process for removal of sulfur compounds from fuel gases

    Moore, Raymond H.; Stegen, Gary E.

    1978-01-01

    Fuel gases such as those produced in the gasification of coal are stripped of sulfur compounds and particulate matter by contact with molten metal salt. The fuel gas and salt are intimately mixed by passage through a venturi or other constriction in which the fuel gas entrains the molten salt as dispersed droplets to a gas-liquid separator. The separated molten salt is divided into a major and a minor flow portion with the minor flow portion passing on to a regenerator in which it is contacted with steam and carbon dioxide as strip gas to remove sulfur compounds. The strip gas is further processed to recover sulfur. The depleted, minor flow portion of salt is passed again into contact with the fuel gas for further sulfur removal from the gas. The sulfur depleted, fuel gas then flows through a solid absorbent for removal of salt droplets. The minor flow portion of the molten salt is then recombined with the major flow portion for feed to the venturi.

  14. Update and evaluation of decay data for spent nuclear fuel analyses

    Simeonov Teodosi

    2017-01-01

    Full Text Available Studsvik’s approach to spent nuclear fuel analyses combines isotopic concentrations and multi-group cross-sections, calculated by the CASMO5 or HELIOS2 lattice transport codes, with core irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the code SNF to predict spent nuclear fuel characteristics. Recent advances in the generation procedure for the SNF decay data are presented. The SNF decay data includes basic data, such as decay constants, atomic masses and nuclide transmutation chains; radiation emission spectra for photons from radioactive decay, alpha-n reactions, bremsstrahlung, and spontaneous fission, electrons and alpha particles from radioactive decay, and neutrons from radioactive decay, spontaneous fission, and alpha-n reactions; decay heat production; and electro-atomic interaction data for bremsstrahlung production. These data are compiled from fundamental (ENDF, ENSDF, TENDL and processed (ESTAR sources for nearly 3700 nuclides. A rigorous evaluation procedure of internal consistency checks and comparisons to measurements and benchmarks, and code-to-code verifications is performed at the individual isotope level and using integral characteristics on a fuel assembly level (e.g., decay heat, radioactivity, neutron and gamma sources. Significant challenges are presented by the scope and complexity of the data processing, a dearth of relevant detailed measurements, and reliance on theoretical models for some data.

  15. Update and evaluation of decay data for spent nuclear fuel analyses

    Simeonov, Teodosi; Wemple, Charles

    2017-09-01

    Studsvik's approach to spent nuclear fuel analyses combines isotopic concentrations and multi-group cross-sections, calculated by the CASMO5 or HELIOS2 lattice transport codes, with core irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the code SNF to predict spent nuclear fuel characteristics. Recent advances in the generation procedure for the SNF decay data are presented. The SNF decay data includes basic data, such as decay constants, atomic masses and nuclide transmutation chains; radiation emission spectra for photons from radioactive decay, alpha-n reactions, bremsstrahlung, and spontaneous fission, electrons and alpha particles from radioactive decay, and neutrons from radioactive decay, spontaneous fission, and alpha-n reactions; decay heat production; and electro-atomic interaction data for bremsstrahlung production. These data are compiled from fundamental (ENDF, ENSDF, TENDL) and processed (ESTAR) sources for nearly 3700 nuclides. A rigorous evaluation procedure of internal consistency checks and comparisons to measurements and benchmarks, and code-to-code verifications is performed at the individual isotope level and using integral characteristics on a fuel assembly level (e.g., decay heat, radioactivity, neutron and gamma sources). Significant challenges are presented by the scope and complexity of the data processing, a dearth of relevant detailed measurements, and reliance on theoretical models for some data.

  16. Fuel processing for PEM fuel cells: transport and kinetic issues of system design

    Zalc, J. M.; Löffler, D. G.

    In light of the distribution and storage issues associated with hydrogen, efficient on-board fuel processing will be a significant factor in the implementation of PEM fuel cells for automotive applications. Here, we apply basic chemical engineering principles to gain insight into the factors that limit performance in each component of a fuel processor. A system consisting of a plate reactor steam reformer, water-gas shift unit, and preferential oxidation reactor is used as a case study. It is found that for a steam reformer based on catalyst-coated foils, mass transfer from the bulk gas to the catalyst surface is the limiting process. The water-gas shift reactor is expected to be the largest component of the fuel processor and is limited by intrinsic catalyst activity, while a successful preferential oxidation unit depends on strict temperature control in order to minimize parasitic hydrogen oxidation. This stepwise approach of sequentially eliminating rate-limiting processes can be used to identify possible means of performance enhancement in a broad range of applications.

  17. Summary of Preliminary Criticality Analysis for Peach Bottom Fuel in the DOE Standardized Spent Nuclear Fuel Canister

    Henrikson, D.J.

    1999-01-01

    The Department of Energy's (DOE's) National Spent Nuclear Fuel Program is developing a standardized set of canisters for DOE spent nuclear fuel (SNF). These canisters will be used for DOE SNF handling, interim storage, transportation, and disposal in the national repository. Several fuels are being examined in conjunction with the DOE SNF canisters. This report summarizes the preliminary criticality safety analysis that addresses general fissile loading limits for Peach Bottom graphite fuel in the DOE SNF canister. The canister is considered both alone and inside the 5-HLW/DOE Long Spent Fuel Co-disposal Waste Package, and in intact and degraded conditions. Results are appropriate for a single DOE SNF canister. Specific facilities, equipment, canister internal structures, and scenarios for handling, storage, and transportation have not yet been defined and are not evaluated in this analysis. The analysis assumes that the DOE SNF canister is designed so that it maintains reasonable geometric integrity. Parameters important to the results are the canister outer diameter, inner diameter, and wall thickness. These parameters are assumed to have nominal dimensions of 45.7-cm (18.0-in.), 43.815-cm (17.25-in), and 0.953-cm (0.375-in.), respectively. Based on the analysis results, the recommended fissile loading for the DOE SNF canister is 13 Peach Bottom fuel elements if no internal steel is present, and 15 Peach Bottom fuel elements if credit is taken for internal steel

  18. Solvent extraction process development for high plutonium fuel cycles

    Anil Kumar, R; Selvaraj, P G; Natarajan, R; Raman, V R [Reprocessing Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1994-06-01

    The purification of high plutonium bearing irradiated fuels using 30% TBP in dodecane diluent requires precise determination of concentration profiles during steady state, transient and process upset conditions. Mathematical models have been developed and a computer code is in use for determining Pu-U concentration profiles in a solvent extraction equipment in a typical reprocessing plant. The process parameters have been optimised for recovery of U and Pu and decontamination from the fission products. This computer code is used to analyse the extraction flow sheets of fuels of two typical Pu-U compositions encountered in Indian fast breeder programme. The analysis include the effect of uncertainty in equilibrium condition prediction by the model and the variation of flows of streams during plant operation. The studies highlight the margin available to avoid second organic phase formation and adjustments required in the process flowsheet. (author). 7 refs., 7 figs., 2 tabs.

  19. Process Investigation for Conversion of MSW into Liquid Fuel

    Javed, M.T.; Jafri, U.A.; Chugtai, I.R.

    2010-01-01

    An investigation was conducted on pyrolysis technology to convert the municipal solid waste into liquid fuel. The investigation includes the development of the experimental setup for this process and its future prospects in Pakistan. A pyrolysis process is under consideration for many years for the production of synthetic fuel oils from organic solid waste. The system comprises of pyrolysis reactor, condenser for condensable gas, gas holder (for non- condensable gas). The feedstock used in the pyrolysis reactor is the municipal solid waste (includes kitchen waste, papers etc) in fine mesh size i.e. 2.5 - 3.0 mm. The residue obtained were mainly tar (pyrolytic oil), pyrogas (non - condensable gases) and ash, which shows that process has a potential for the treatment of the municipal solid waste and is a good technology for resource recover. (author)

  20. Used nuclear fuel separations process simulation and testing

    Pereira, C.; Krebs, J.F.; Copple, J.M.; Frey, K.E.; Maggos, L.E.; Figueroa, J.; Willit, J.L.; Papadias, D.D.

    2013-01-01

    Recent efforts in separations process simulation at Argonne have expanded from the traditional focus on solvent extraction flowsheet design in order to capture process dynamics and to simulate other components, processing and systems of a used nuclear fuel reprocessing plant. For example, the Argonne Model for Universal Solvent Extraction (AMUSE) code has been enhanced to make it both more portable and more readily extensible. Moving away from a spreadsheet environment makes the addition of new species and processes simpler for the expert user, which should enable more rapid implementation of chemical models that simulate evolving processes. The dyAMUSE (dynamic AMUSE) version allows the simulation of transient behavior across an extractor. Electrochemical separations have now been modeled using spreadsheet codes that simulate the electrochemical recycle of fast reactor fuel. The user can follow the evolution of the salt, products, and waste compositions in the electro-refiner, cathode processors, and drawdown as a function of fuel batches treated. To further expand capabilities in integrating multiple unit operations, a platform for linking mathematical models representing the different operations that comprise a reprocessing facility was adapted to enable systems-level analysis and optimization of facility functions. (authors)

  1. Used nuclear fuel separations process simulation and testing

    Pereira, C.; Krebs, J.F.; Copple, J.M.; Frey, K.E.; Maggos, L.E.; Figueroa, J.; Willit, J.L.; Papadias, D.D. [Argonne National Laboratory: 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2013-07-01

    Recent efforts in separations process simulation at Argonne have expanded from the traditional focus on solvent extraction flowsheet design in order to capture process dynamics and to simulate other components, processing and systems of a used nuclear fuel reprocessing plant. For example, the Argonne Model for Universal Solvent Extraction (AMUSE) code has been enhanced to make it both more portable and more readily extensible. Moving away from a spreadsheet environment makes the addition of new species and processes simpler for the expert user, which should enable more rapid implementation of chemical models that simulate evolving processes. The dyAMUSE (dynamic AMUSE) version allows the simulation of transient behavior across an extractor. Electrochemical separations have now been modeled using spreadsheet codes that simulate the electrochemical recycle of fast reactor fuel. The user can follow the evolution of the salt, products, and waste compositions in the electro-refiner, cathode processors, and drawdown as a function of fuel batches treated. To further expand capabilities in integrating multiple unit operations, a platform for linking mathematical models representing the different operations that comprise a reprocessing facility was adapted to enable systems-level analysis and optimization of facility functions. (authors)

  2. Multi Canister Overpack (MCO) Closure Welding Process Parameter Development and Qualification

    CANNELL, G.R.

    2003-01-01

    One of the Department of Energy's (DOE) top priorities at the Hanford Site (southeastern Washington state), is the processing of more than 2,000 tons of spent nuclear fuel (SNF) into large stainless steel containers called Multi-Canister Overpacks (MCO). Packaging into MCO's will assist in the safe and economic disposition of SNF and greatly reduce risk to the environment. Packaged fuel will be removed from close proximity to the Columbia River to a more suitable area of the site where it will be stored on an interim basis. Eventually, the fuel will be transferred to the federal geologic repository for long-term storage. One of the key elements in the SNF process is final closure of the MCO by welding. Fuel is loaded into the MCO (approximately 2 ft. in diameter and 13 ft. long) and a heavy shield plug inserted into the top, creating a mechanical seal. The plug contains several process ports for various operations, including vacuum drying and inert-gas backfilling of the packaged fuel. When fully processed, the Canister Cover Assembly (CCA) is placed over the shield plug and final closure made by welding. The following describes the effort to develop and qualify the root-pass technique associated with the MCO final closure weld

  3. Concerning permission of change in nuclear fuel processing business of Japan Nuclear Fuel Co. , Ltd

    1988-12-01

    In response to an inquiry on the title issue received on Jun. 17, 1988, the Nuclear Safety Commission made a study and submitted the findings to the Prime Minister on Jul. 21, 1988. The study was intended to determine the conformity of the permission to the applicable criteria specified in laws relating to control of nuclear material, nuclear fuel and nuclear reactor. The proposed modification plan included changes in the facilities in the No.1 processing building and changes in processing methods which were required to perform processing of blanket fuel assemblies for fast breeder reactor. It also included changes in the facilities in the No.2 building which were required to improve the processes. The safety study covered the anti-earthquake performance, fire/explosion prevention, criticality control, containment performance, radioactive waste disposal, and other major safety issues. Other investigations included exposure dose evaluation and accident analysis. Study results were examined on the basis of the Basic Guidelines for Nuclear Fuel Facilities Safety Review and the Uranium Processing Safety Review Guidelines. It was concluded that the modifications would not have adverse effect on the safety of the facilities. (Nogami, K.).

  4. Concerning permission of change in nuclear fuel processing business of Japan Nuclear Fuel Co., Ltd

    1988-01-01

    In response to an inquiry on the title issue received on Jun. 17, 1988, the Nuclear Safety Commission made a study and submitted the findings to the Prime Minister on Jul. 21, 1988. The study was intended to determine the conformity of the permission to the applicable criteria specified in laws relating to control of nuclear material, nuclear fuel and nuclear reactor. The proposed modification plan included changes in the facilities in the No.1 processing building and changes in processing methods which were required to perform processing of blanket fuel assemblies for fast breeder reactor. It also included changes in the facilities in the No.2 building which were required to improve the processes. The safety study covered the anti-earthquake performance, fire/explosion prevention, criticality control, containment performance, radioactive waste disposal, and other major safety issues. Other investigations included exposure dose evaluation and accident analysis. Study results were examined on the basis of the Basic Guidelines for Nuclear Fuel Facilities Safety Review and the Uranium Processing Safety Review Guidelines. It was concluded that the modifications would not have adverse effect on the safety of the facilities. (Nogami, K.)

  5. Pyrometallurgical processing of Integral Fast Reactor metal fuels

    Battles, J.E.; Miller, W.E.; Gay, E.C.

    1991-01-01

    The pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor is now in an advanced state of development. This process involves electrorefining spent fuel with a cadmium anode, solid and liquid cathodes, and a molten salt electrolyte (LiCl-KCl) at 500 degrees C. The initial process feasibility and flowsheet verification studies have been conducted in a laboratory-scale electrorefiner. Based on these studies, a dual cathode approach has been adopted, where uranium is recovered on a solid cathode mandrel and uranium-plutonium is recovered in a liquid cadmium cathode. Consolidation and purification (salt and cadmium removal) of uranium and uranium-plutonium products from the electrorefiner have been successful. The process is being developed with the aid of an engineering-scale electrorefiner, which has been successfully operated for more than three years. In this electrorefiner, uranium has been electrotransported from the cadmium anode to a solid cathode in 10 kg quantities. Also, anodic dissolution of 10 kg batches of chopped, simulated fuel (U--10% Zr) has been demonstrated. Development of the liquid cadmium cathode for recovering uranium-plutonium is under way

  6. Towards Extrusion of Ionomers to Process Fuel Cell Membranes

    Jean-Yves Sanchez

    2011-07-01

    Full Text Available While Proton Exchange Membrane Fuel Cell (PEMFC membranes are currently prepared by film casting, this paper demonstrates the feasibility of extrusion, a solvent-free alternative process. Thanks to water-soluble process-aid plasticizers, duly selected, it was possible to extrude acidic and alkaline polysulfone ionomers. Additionally, the feasibility to extrude composites was demonstrated. The impact of the plasticizers on the melt viscosity was investigated. Following the extrusion, the plasticizers were fully removed in water. The extrusion was found to impact neither on the ionomer chains, nor on the performances of the membrane. This environmentally friendly process was successfully validated for a variety of high performance ionomers.

  7. Spent nuclear fuel storage: Legal, technical and political considerations

    Blake, E.L. Jr.; Buren, M.A.

    1994-01-01

    In 1982, Congress enacted the Nuclear Waste Policy Act (NWPA), assigning responsibility to the Department of Energy (DOE) for the development and implementation of a comprehensive national nuclear waste management program. The NWPA makes clear that the generators and owners of commercially-generated spent nuclear fuel (SNF) have the primary responsibility to provide for, and pay the costs of, the interim storage of such SNF until it is accepted by the DOE under the provisions of the NWPA. The shift in responsibility was expected to begin in 1998, the date specified in the NWPA and the DOE's contracts with the utilities, at which time the NWPA anticipated commencement of operations of a geologic repository and/or a monitored retrievable storage facility (MRS). Unfortunately, despite a mid-course correction to the NWPA mandated by Congress in 1987 in an effort to streamline and accelerate the program, DOE is way behind schedule. DOE's last published program schedule indicates the commencement of repository operations in 2010, a date many feel is overly optimistic. In repeated statements during the early 1990s, DOE sought to reassure utility companies and their regulatory commissions that it could still commence SNF acceptance in 1998 for storage at an MRS if such a facility were sited through a voluntary process by the end of 1992. That date has now come and gone. Although DOE is still nominally seeking a voluntary MRS host jurisdiction, the prospects for MRS operation by 1998 are dim. Putting aside for the moment the question of DOE's ability to bring the repository on line, the immediate problem facing domestic utilities is the need to augment their onsite SNF storage capacity. In addition to providing a brief overview of the Federal independent spent fuel storage installation (ISFSI) licensing process, the author provides some insight of what the real issues are in ISFSI licensing

  8. Modular, High-Volume Fuel Cell Leak-Test Suite and Process

    Ru Chen; Ian Kaye

    2012-03-12

    Fuel cell stacks are typically hand-assembled and tested. As a result the manufacturing process is labor-intensive and time-consuming. The fluid leakage in fuel cell stacks may reduce fuel cell performance, damage fuel cell stack, or even cause fire and become a safety hazard. Leak check is a critical step in the fuel cell stack manufacturing. The fuel cell industry is in need of fuel cell leak-test processes and equipment that is automatic, robust, and high throughput. The equipment should reduce fuel cell manufacturing cost.

  9. Radiation Monitoring System in Advanced Spent Fuel Conditioning Process Facility

    You, Gil Sung; Kook, D. H.; Choung, W. M.; Ku, J. H.; Cho, I. J.; You, G. S.; Kwon, K. C.; Lee, W. K.; Lee, E. P

    2006-09-15

    The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO{sub 2} into U-metal. For demonstration of this process, {alpha}-{gamma} type new hot cell was built in the IMEF basement . To secure against radiation hazard, this facility needs radiation monitoring system which will observe the entire operating area before the hot cell and service area at back of it. This system consists of 7 parts; Area Monitor for {gamma}-ray, Room Air Monitor for particulate and iodine in both area, Hot cell Monitor for hot cell inside high radiation and rear door interlock, Duct Monitor for particulate of outlet ventilation, Iodine Monitor for iodine of outlet duct, CCTV for watching workers and material movement, Server for management of whole monitoring system. After installation and test of this, radiation monitoring system will be expected to assist the successful ACP demonstration.

  10. Radiation Monitoring System in Advanced Spent Fuel Conditioning Process Facility

    You, Gil Sung; Kook, D. H.; Choung, W. M.; Ku, J. H.; Cho, I. J.; You, G. S.; Kwon, K. C.; Lee, W. K.; Lee, E. P.

    2006-09-01

    The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO 2 into U-metal. For demonstration of this process, α-γ type new hot cell was built in the IMEF basement . To secure against radiation hazard, this facility needs radiation monitoring system which will observe the entire operating area before the hot cell and service area at back of it. This system consists of 7 parts; Area Monitor for γ-ray, Room Air Monitor for particulate and iodine in both area, Hot cell Monitor for hot cell inside high radiation and rear door interlock, Duct Monitor for particulate of outlet ventilation, Iodine Monitor for iodine of outlet duct, CCTV for watching workers and material movement, Server for management of whole monitoring system. After installation and test of this, radiation monitoring system will be expected to assist the successful ACP demonstration

  11. Process control of an HTGR fuel reprocessing cold pilot plant

    Rode, J.S.

    1976-10-01

    Development of engineering-scale systems for a large-scale HTGR fuel reprocessing demonstration facility is currently underway in a cold pilot plant. These systems include two fluidized-bed burners, which remove the graphite (carbon) matrix from the crushed HTGR fuel by high temperature (900 0 C) oxidation. The burners are controlled by a digital process controller with an all analog input/output interface which has been in use since March, 1976. The advantages of such a control system to a pilot plant operation can be summarized as follows: (1) Control loop functions and configurations can be changed easily; (2) control constants, alarm limits, output limits, and scaling constants can be changed easily; (3) calculation of data and/or interface with a computerized information retrieval system during operation are available; (4) diagnosis of process control problems is facilitated; and (5) control panel/room space is saved

  12. Spent Nuclear Fuel Alternative Technology Decision Analysis

    Shedrow, C.B.

    1999-01-01

    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

  13. Spent Nuclear Fuel Alternative Technology Decision Analysis

    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.

  14. Improvement of the Zircaloy fuel can manufacturing process

    1986-01-01

    The following work has been performed in order to ensure more reliable supply of start material for the manufacture of Zy-2 and Zy-4 fuel cans, and to improve the processing techniques and product quality: 1) Two complete production campaigns with the ingot suppliers Western Zirconium and Ugine Aciers. 2) Development of new ingot dimensions (rolling tests). 3) Development of a mechanized washing and cleansing procedure. 4) Development of a new abrasive treatment technique (wet sand blasting). (orig./HP) [de

  15. Vertical Drop of the Naval SNF Long Waste Package On Unyielding Surface

    S. Mastilovic

    2006-01-01

    The purpose of this calculation is to determine the structural response of a Naval SNF (Spent Nuclear Fuel) Long Waste Package (WP) subjected to 2 m-vertical drop on unyielding surface (US). The scope of this document is limited to reporting the calculation results in terms of maximum stress intensities. This calculation is associated with the waste package design; calculation is performed by the Waste Package Design group. AP-3.12Q, Revision 0, ICN 0, Calculations, is used to perform the calculation and develop the document. The finite element calculation is performed by using the commercially available ANSYS Version (V) 5.4 finite element code. The result of this calculation is provided in terms of maximum stress intensities

  16. Downregulation of SWI/SNF chromatin remodeling factor subunits modulates cisplatin cytotoxicity

    Kothandapani, Anbarasi; Gopalakrishnan, Kathirvel; Kahali, Bhaskar; Reisman, David; Patrick, Steve M.

    2012-01-01

    Chromatin remodeling complex SWI/SNF plays important roles in many cellular processes including transcription, proliferation, differentiation and DNA repair. In this report, we investigated the role of SWI/SNF catalytic subunits Brg1 and Brm in the cellular response to cisplatin in lung cancer and head/neck cancer cells. Stable knockdown of Brg1 and Brm enhanced cellular sensitivity to cisplatin. Repair kinetics of cisplatin DNA adducts revealed that downregulation of Brg1 and Brm impeded the repair of both intrastrand adducts and interstrand crosslinks (ICLs). Cisplatin ICL-induced DNA double strand break repair was also decreased in Brg1 and Brm depleted cells. Altered checkpoint activation with enhanced apoptosis as well as impaired chromatin relaxation was observed in Brg1 and Brm deficient cells. Downregulation of Brg1 and Brm did not affect the recruitment of DNA damage recognition factor XPC to cisplatin DNA lesions, but affected ERCC1 recruitment, which is involved in the later stages of DNA repair. Based on these results, we propose that SWI/SNF chromatin remodeling complex modulates cisplatin cytotoxicity by facilitating efficient repair of the cisplatin DNA lesions. -- Highlights: ► Stable knockdown of Brg1 and Brm enhances cellular sensitivity to cisplatin. ► Downregulation of Brg1 and Brm impedes the repair of cisplatin intrastrand adducts and interstrand crosslinks. ► Brg1 and Brm deficiency results in impaired chromatin relaxation, altered checkpoint activation as well as enhanced apoptosis. ► Downregulation of Brg1 and Brm affects recruitment of ERCC1, but not XPC to cisplatin DNA lesions.

  17. The SNF2-family member Fun30 promotes gene silencing in heterochromatic loci.

    Ana Neves-Costa

    2009-12-01

    Full Text Available Chromatin regulates many key processes in the nucleus by controlling access to the underlying DNA. SNF2-like factors are ATP-driven enzymes that play key roles in the dynamics of chromatin by remodelling nucleosomes and other nucleoprotein complexes. Even simple eukaryotes such as yeast contain members of several subfamilies of SNF2-like factors. The FUN30/ETL1 subfamily of SNF2 remodellers is conserved from yeasts to humans, but is poorly characterized. We show that the deletion of FUN30 leads to sensitivity to the topoisomerase I poison camptothecin and to severe cell cycle progression defects when the Orc5 subunit is mutated. We demonstrate a role of FUN30 in promoting silencing in the heterochromatin-like mating type locus HMR, telomeres and the rDNA repeats. Chromatin immunoprecipitation experiments demonstrate that Fun30 binds at the boundary element of the silent HMR and within the silent HMR. Mapping of nucleosomes in vivo using micrococcal nuclease demonstrates that deletion of FUN30 leads to changes of the chromatin structure at the boundary element. A point mutation in the ATP-binding site abrogates the silencing function of Fun30 as well as its toxicity upon overexpression, indicating that the ATPase activity is essential for these roles of Fun30. We identify by amino acid sequence analysis a putative CUE motif as a feature of FUN30/ETL1 factors and show that this motif assists Fun30 activity. Our work suggests that Fun30 is directly involved in silencing by regulating the chromatin structure within or around silent loci.

  18. Closing nuclear fuel cycle with fast reactors: problems and prospects

    Shadrin, A.; Dvoeglazov, K.; Ivanov, V. [Bochvar Institute - VNIINM, Moscow (Russian Federation)

    2013-07-01

    The closed nuclear fuel cycle (CNFC) with fast reactors (FR) is the most promising way of nuclear energetics development because it prevents spent nuclear fuel (SNF) accumulation and minimizes radwaste volume due to minor actinides (MA) transmutation. CNFC with FR requires the elaboration of safety, environmentally acceptable and economically effective methods of treatment of SNF with high burn-up and low cooling time. The up-to-date industrially implemented SNF reprocessing technologies based on hydrometallurgical methods are not suitable for the reprocessing of SNF with high burn-up and low cooling time. The alternative dry methods (such as electrorefining in molten salts or fluoride technologies) applicable for such SNF reprocessing have not found implementation at industrial scale. So the cost of SNF reprocessing by means of dry technologies can hardly be estimated. Another problem of dry technologies is the recovery of fissionable materials pure enough for dense fuel fabrication. A combination of technical solutions performed with hydrometallurgical and dry technologies (pyro-technology) is proposed and it appears to be a promising way for the elaboration of economically, ecologically and socially accepted technology of FR SNF management. This paper deals with discussion of main principle of dry and aqueous operations combination that probably would provide safety and economic efficiency of the FR SNF reprocessing. (authors)

  19. Spent nuclear fuel project multi-year work plan WBS number 1.4.1

    Wells, J.L.

    1997-01-01

    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

  20. Process for the fabrication of nuclear fuel oxide pellets

    Francois, Bernard; Paradis, Yves.

    1977-01-01

    Process for the fabrication of nuclear fuel oxide pellets of the type for which particles charged with an organic binder -selected from the group that includes polyvinyl alcohol, carboxymethyl cellulose, polyvinyl compounds and methyl cellulose- are prepared from a powder of such an oxide, for instance uranium dioxide. These particles are then compressed into pellets which are then sintered. Under this process the binder charged particles are prepared by stirring the powder with a gas, spraying on to the stirred powder a solution or a suspension in a liquid of this organic binder in order to obtain these particles and then drying the particles so obtained with this gas [fr

  1. Smelting Associated with the Advanced Spent Fuel Conditioning Process

    Hur, J-M.; Jeong, M-S.; Lee, W-K.; Cho, S-H.; Seo, C-S.; Park, S-W.

    2004-01-01

    The smelting process associated with the advanced spent fuel conditioning process (ACP) of Korea Atomic Energy Research Institute was studied by using surrogate materials. Considering the vaporization behaviors of input materials, the operation procedure of smelting was set up as (1) removal of residual salts, (2) melting of metal powder, and (3) removal of dross from a metal ingot. The behaviors of porous MgO crucible during smelting were tested and the chemical stability of MgO in the salt-being atmosphere was confirmed

  2. Purification and characterization of the three Snf1-activating kinases of Saccharomyces cerevisiae.

    Elbing, Karin; McCartney, Rhonda R; Schmidt, Martin C

    2006-02-01

    Members of the Snf1/AMPK family of protein kinases are activated by distinct upstream kinases that phosphorylate a conserved threonine residue in the Snf1/AMPK activation loop. Recently, the identities of the Snf1- and AMPK-activating kinases have been determined. Here we describe the purification and characterization of the three Snf1-activating kinases of Saccharomyces cerevisiae. The identities of proteins associated with the Snf1-activating kinases were determined by peptide mass fingerprinting. These kinases, Sak1, Tos3 and Elm2 do not appear to require the presence of additional subunits for activity. Sak1 and Snf1 co-purify and co-elute in size exclusion chromatography, demonstrating that these two proteins form a stable complex. The Snf1-activating kinases phosphorylate the activation loop threonine of Snf1 in vitro with great specificity and are able to do so in the absence of beta and gamma subunits of the Snf1 heterotrimer. Finally, we showed that the Snf1 kinase domain isolated from bacteria as a GST fusion protein can be activated in vitro and shows substrate specificity in the absence of its beta and gamma subunits.

  3. Status of Progress Made Toward Safety Analysis and Technical Site Evaluations for DOE Managed HLW and SNF.

    Sevougian, S. David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Stein, Emily [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gross, Michael B [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hammond, Glenn Edward [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Frederick, Jennifer M [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mariner, Paul [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-11-01

    The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) is conducting research and development (R&D) on generic deep geologic disposal systems (i.e., repositories). This report describes specific activities in FY 2016 associated with the development of a Defense Waste Repository (DWR)a for the permanent disposal of a portion of the HLW and SNF derived from national defense and research and development (R&D) activities of the DOE.

  4. Structure and novel functional mechanism of Drosophila SNF in sex-lethal splicing.

    Jicheng Hu

    Full Text Available Sans-fille (SNF is the Drosophila homologue of mammalian general splicing factors U1A and U2B'', and it is essential in Drosophila sex determination. We found that, besides its ability to bind U1 snRNA, SNF can also bind polyuridine RNA tracts flanking the male-specific exon of the master switch gene Sex-lethal (Sxl pre-mRNA specifically, similar to Sex-lethal protein (SXL. The polyuridine RNA binding enables SNF directly inhibit Sxl exon 3 splicing, as the dominant negative mutant SNF(1621 binds U1 snRNA but not polyuridine RNA. Unlike U1A, both RNA recognition motifs (RRMs of SNF can recognize polyuridine RNA tracts independently, even though SNF and U1A share very high sequence identity and overall structure similarity. As SNF RRM1 tends to self-associate on the opposite side of the RNA binding surface, it is possible for SNF to bridge the formation of super-complexes between two introns flanking Sxl exon 3 or between a intron and U1 snRNP, which serves the molecular basis for SNF to directly regulate Sxl splicing. Taken together, a new functional model for SNF in Drosophila sex determination is proposed. The key of the new model is that SXL and SNF function similarly in promoting Sxl male-specific exon skipping with SNF being an auxiliary or backup to SXL, and it is the combined dose of SXL and SNF governs Drosophila sex determination.

  5. U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

    Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peterson, Joshua L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decay heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.

  6. A survey of processes for producing hydrogen fuel from different sources for automotive-propulsion fuel cells

    Brown, L.F.

    1996-03-01

    Seven common fuels are compared for their utility as hydrogen sources for proton-exchange-membrane fuel cells used in automotive propulsion. Methanol, natural gas, gasoline, diesel fuel, aviation jet fuel, ethanol, and hydrogen are the fuels considered. Except for the steam reforming of methanol and using pure hydrogen, all processes for generating hydrogen from these fuels require temperatures over 1000 K at some point. With the same two exceptions, all processes require water-gas shift reactors of significant size. All processes require low-sulfur or zero-sulfur fuels, and this may add cost to some of them. Fuels produced by steam reforming contain {approximately}70-80% hydrogen, those by partial oxidation {approximately}35-45%. The lower percentages may adversely affect cell performance. Theoretical input energies do not differ markedly among the various processes for generating hydrogen from organic-chemical fuels. Pure hydrogen has severe distribution and storage problems. As a result, the steam reforming of methanol is the leading candidate process for on-board generation of hydrogen for automotive propulsion. If methanol unavailability or a high price demands an alternative process, steam reforming appears preferable to partial oxidation for this purpose.

  7. Technical and regulatory review of the Rover nuclear fuel process for use on Fort St. Vrain fuel

    Hertzler, T.

    1993-02-01

    This report describes the results of an analysis for processing and final disposal of Fort St. Vrain (FSV) irradiated fuel in Rover-type equipment or technologies. This analysis includes an evaluation of the current Rover equipment status and the applicability of this technology in processing FSV fuel. The analyses are based on the physical characteristics of the FSV fuel and processing capabilities of the Rover equipment. Alternate FSV fuel disposal options are also considered including fuel-rod removal from the block, disposal of the empty block, or disposal of the entire fuel-containing block. The results of these analyses document that the current Rover hardware is not operable for any purpose, and any effort to restart this hardware will require extensive modifications and re-evaluation. However, various aspects of the Rover technology, such as the successful fluid-bed burner design, can be applied with modification to FSV fuel processing. The current regulatory climate and technical knowledge are not adequately defined to allow a complete analysis and conclusion with respect to the disposal of intact fuel blocks with or without the fuel rods removed. The primary unknowns include the various aspects of fuel-rod removal from the block, concentration of radionuclides remaining in the graphite block after rod removal, and acceptability of carbon in the form of graphite in a high level waste repository

  8. Quality control of CANDU6 fuel element in fabrication process

    Li Yinxie; Zhang Jie

    2012-01-01

    To enhance the fine control over all aspects of the production process, improve product quality, fuel element fabrication process for CANDU6 quality process control activities carried out by professional technical and management technology combined mode, the quality of the fuel elements formed around CANDU6 weak links - - end plug , and brazing processes and procedures associated with this aspect of strict control, in improving staff quality consciousness, strengthening equipment maintenance, improved tooling, fixtures, optimization process test, strengthen supervision, fine inspection operations, timely delivery carry out aspects of the quality of information and concerns the production environment, etc., to find the problem from the improvement of product quality and factors affecting the source, and resolved to form the active control, comprehensive and systematic analysis of the problem of the quality management concepts, effectively reducing the end plug weld microstructure after the failure times and number of defects zirconium alloys brazed, improved product quality, and created economic benefits expressly provided, while staff quality consciousness and attention to detail, collaboration department, communication has been greatly improved and achieved very good management effectiveness. (authors)

  9. Iodine-129 in aquatic organisms near nuclear fuels processing plants

    Watson, D.G.

    1975-04-01

    Concentrations of 129 I in two aquatic habitats near nuclear fuel processing plants were highest in algae and crustaceans. These two forms may be useful in future monitoring of 129 I. There is some indication of an increase in atom ratios and specific activity in aquatic organisms over that in water and sediments. Additional measurements should be made to verify this conclusion. Efforts should continue to measure the possible long term build-up of 129 I in aquatic environments receiving effluents from fuels reprocessing plants. Even at very low rates of release to the environment, the long physical half-life of 129 I creates the potential for build-up of this nuclide to significant levels. (U.S.)

  10. Use of Hansen Solubility Parameters in Fuel Treatment Processes

    2014-03-17

    Clearance # Considerations for Rocket Fuel Objective: Utilize liquid/liquid extraction process to improve performance, increase availability, and...1/4)(H1 - H0)2 - (D2 – D0)2 - (1/4) (P2 - P0)2 - (1/4)(H2 - H0)2 ] + RT ln (V1/ V2 ) K = C0,2 / CO,1 Partition coefficient RT ln K = V0( D1...02 – D2-02 ) + RT ln (V1/ V2 ) Di-0 is the distance in “solubility parameter space” between liquid i and impurity 0. For reference, phase 1 = fuel

  11. U.S. report on fuel performance and technology

    Cook, T [Department of Energy, Washington, DC (United States). Office of Engineering and Technology Development

    1997-12-01

    The report reviews the following aspects of fuel performance and technology: increased demand on fuel performance;improved fuel failure rate; operating fuel cycles; capacity factor for US nuclear electric generating plants; potential reduction of SNF due to improved fuel burnup.

  12. Hydrocarbon fuel processing of micro solid oxide fuel cell systems[Dissertation 17455

    Stutz, M. J.

    2007-07-01

    The scope of this thesis is the numerical and experimental investigation of the fuel processing of a micro solid oxide fuel cell (SOFC) running on hydrocarbon fuel. The goal is to enhance the overall system efficiency by optimization of the reforming process in the steady state and the improvement of the start-up process. Micro SOFC are a potential alternative to the currently used batteries in portable devices. Liquid butane in a cartridge could be the energy source. This dissertation is focused on the fuel processing of the system, namely the reforming and post-combusting processes. The reformer converts the hydrocarbon fuel to a hydrogen rich gas that can be utilized by the SOFC. The post-combustor depletes the toxic and/or explosive gases before leaving the exhaust. Chapter One presents a short introduction to the field of hydrocarbon fuel processing in micro solid oxide fuel cell systems, the next three chapters deal with computational modeling of the transport phenomena inside a micro-reformer, which leads to a better understanding of the chemistry and the physics therein, hence progress in the design and operation parameters. The experimental part (i.e. Chapter Five) of this thesis focuses on the feasibility of a novel hybrid start-up method of a fuel cell system that employs existing components as an additional heat source. In Chapter Two the effect of wall heat conduction on the syngas (hydrogen and carbon monoxide) production of a micro-reformer, representing micro-fabricated channels or monoliths, is investigated. Methane is used as a model hydrocarbon fuel since its heterogeneous reaction path on rhodium is known and validated. The simulations demonstrate that the axial wall conduction strongly influences the performance of the micro-reformer and should not be neglected without a careful a priori investigation of its impact. Methane conversion and hydrogen yield are strongly dependent of the wall inner surface temperature, which is influenced by the

  13. U-10Mo Baseline Fuel Fabrication Process Description

    Hubbard, Lance R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Arendt, Christina L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Dye, Daniel F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Clayton, Christopher K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lerchen, Megan E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lombardo, Nicholas J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zacher, Alan H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-09-27

    This document provides a description of the U.S. High Power Research Reactor (USHPRR) low-enriched uranium (LEU) fuel fabrication process. This document is intended to be used in conjunction with the baseline process flow diagram (PFD) presented in Appendix A. The baseline PFD is used to document the fabrication process, communicate gaps in technology or manufacturing capabilities, convey alternatives under consideration, and as the basis for a dynamic simulation model of the fabrication process. The simulation model allows for the assessment of production rates, costs, and manufacturing requirements (manpower, fabrication space, numbers and types of equipment, etc.) throughout the lifecycle of the USHPRR program. This document, along with the accompanying PFD, is updated regularly

  14. Techniques for laser processing, assay, and examination of spent fuel

    Gray, J.H.; Mitchell, R.C.; Rogell, M.L.

    1981-11-01

    Fuel examination studies were performed which have application to interim spent fuel storage. These studies were in three areas, i.e., laser drilling and rewelding demonstration, nondestructive assay techniques survey, and fuel examination techniques survey

  15. Progress in fuel processing for PEMFC systems for transport applications

    Dams, A J; Hayter, P R; Moore, S C

    1998-07-01

    Wellman CJB Limited has been developing fuel processors for PEMFC systems for transport applications using a range of feedstocks. Feedstocks that can be processed to produce a hydrogen-rich gas stream suitable for use with a PEMFC include methanol, gasoline, diesel, LPG, dimethylether, marine diesel and aviation fuel. The basic fuel processor is a steam reformer combined with a selective carbon monoxide oxidation stage. Depending on the nature of the liquid feedstock, other process steps will be required such as vaporizer, desulfurizer, pre-reformer and high and low temperature shift reactors. Work on methanol reforming has been specifically targeted at a PEMFC driven passenger car as part of a multinational European Community JOULE programme called MERCATOX. The objective is to develop and test a compact and fast response methanol reformer and gas clean-up unit to meet a car manufacturer's specification. The method of construction is to coat a methanol reforming catalyst onto one side of a metal corrugated plate. On the other side is a coated combustion catalyst which burns fuel cell off-gas to provide the endothermic heat for the methanol reforming reaction. A number of these plates are assembled in a compact unit ensuring good heat transfer. The gas clean-up unit is similarly constructed with a selective oxidation catalyst on one side and a thermal fluid on the other. Initial tests have indicated a superior performance to conventional packed bed reformers in terms of response and start-up time. Steam reforming of gasoline, diesel, LPG, dimethylether, marine diesel and aviation fuel has been demonstrated on a bench scale (0.5kW). The process steps commence with vaporization (except for LPG), desulfurization (except for dimethylether), prereforming, reforming, low and high temperature shift and selective oxidation. A simple technology demonstrator has shown that a hydrogen-rich mixture (75% hydrogen, 25% carbon dioxide) with less than 2ppm carbon monoxide can be

  16. HYDRIDE-RELATED DEGRADATION OF SNF CLADDING UNDER REPOSITORY CONDITIONS

    McCoy, K.

    2000-01-01

    The purpose and scope of this analysis/model report is to analyze the degradation of commercial spent nuclear fuel (CSNF) cladding under repository conditions by the hydride-related metallurgical processes, such as delayed hydride cracking (DHC), hydride reorientation and hydrogen embrittlement, thereby providing a better understanding of the degradation process and clarifying which aspects of the process are known and which need further evaluation and investigation. The intended use is as an input to a more general analysis of cladding degradation

  17. Thermal characteristics during hydrogen fueling process of type IV cylinder

    Kim, Sung Chan [Department of Fire and Disaster Prevention, Kyungil University, 33, Buhori, Hayang, Kyungsan 712-701 (Korea); Lee, Seung Hoon; Yoon, Kee Bong [Department of Mechanical Engineering, Chung Ang University, 221, Huksuk, Dongjak, Seoul 156-756 (Korea)

    2010-07-15

    Temperature increase during hydrogen fueling process is a significant safety concern of a high pressure hydrogen vessel. Hence, thermal characteristics of a Type IV cylinder during hydrogen filling process need to be understood. In this study, a series of experiments were conducted to quantify the temperature change of the cylinder during hydrogen filling to 35 MPa. Computational fluid dynamics (CFD) analysis was also conducted to simulate the conditions of the experiments. The results predicted by the CFD analysis show reasonable agreement with the experiments and the discrepancy between the CFD results and experimental results decrease with higher initial gas pressures. The upper and the lower parts of the vessel showed a temperature difference in the vertical direction. The upper gas temperature was higher than that of the lower part due to the buoyancy effect in the vessel. The maximum gas temperature was higher than the maximum temperature allowed in the ISO safety code (85 C) for the case in which the vessel was pressurized from 0 MPa to 35 MPa. This work contributes to the understanding of the thermal flow characteristics of the hydrogen filling process and notes that additional efforts should be made to guarantee the safety of a type IV cylinder during the hydrogen fueling process. (author)

  18. Low temperature chemical processing of graphite-clad nuclear fuels

    Pierce, Robert A.

    2017-10-17

    A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

  19. A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

    Sevougian, S. David [Advanced Nuclear Energy Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); MacKinnon, Robert J. [Advanced Nuclear Energy Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Leigh, Christi D. [Defense Waste Management Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Hansen, Frank D. [Geoscience Research and Applications Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States)

    2013-07-01

    The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential

  20. A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

    Sevougian, S. David; MacKinnon, Robert J.; Leigh, Christi D.; Hansen, Frank D.

    2013-01-01

    The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential

  1. Metallographic examination of damaged N reactor spent nuclear fuel element SFEC5,4378

    Marschman, S.C.; Pyecha, T.D.; Abrefah, J.

    1997-08-01

    N-Reactor spent nuclear fuel (SNF) is currently residing underwater in the K Basins at the Hanford site, in Richland, Washington. This report presents results of the metallographic examination of specimens cut from an SNF element (Mark IV-E) with breached cladding. The element had resided in the K-West (KW) Storage Basin for at least 10 years after it was discharged from the N-Reactor. The storage containers in the KW Basin were nominally closed, isolating the SNF elements from the open pool environment. Seven specimens from this Mark IV-E outer fuel element were examined using an optical metallograph. Included were two specimens that had been subjected to a conditioning process recommended by the Independent Technical Assessment Team, two specimens that had been subjected to a conditioning process recommended in the Integrated Process Strategy Report, and three that were in the as-received, as-cut condition. One of the as-received specimens had been cut from the damaged (or breached) end of the element. All other specimens were cut from the undamaged mid-region of the fuel element. The specimens were visually examined to (1) identify uranium hydride inclusions present in the uranium metal fuel, (2) measure the thickness of the oxide layer formed on the uranium edges and assess the apparent integrity and adhesion of the oxide layer, and (3) look for features in the microstructure that might provide an insight into the various corrosion processes that occurred during underwater storage in the KW Basin. These features included, but were not limited to, the integrity of the cladding and the fuel-to-cladding bond, obvious anomalies in the microstructure, excessive pitting or friability of the fuel matrix, and obvious anomalies in the distribution of uranium hydride or uranium carbide inclusions. Also, the observed metallographic features of the conditioned specimens were compared with those of the as-received (unconditioned) specimens. 11 refs., 93 figs., 2 tabs.

  2. Conversion of microalgae to jet fuel: process design and simulation.

    Wang, Hui-Yuan; Bluck, David; Van Wie, Bernard J

    2014-09-01

    Microalgae's aquatic, non-edible, highly genetically modifiable nature and fast growth rate are considered ideal for biomass conversion to liquid fuels providing promise for future shortages in fossil fuels and for reducing greenhouse gas and pollutant emissions from combustion. We demonstrate adaptability of PRO/II software by simulating a microalgae photo-bio-reactor and thermolysis with fixed conversion isothermal reactors adding a heat exchanger for thermolysis. We model a cooling tower and gas floatation with zero-duty flash drums adding solids removal for floatation. Properties data are from PRO/II's thermodynamic data manager. Hydrotreating is analyzed within PRO/II's case study option, made subject to Jet B fuel constraints, and we determine an optimal 6.8% bioleum bypass ratio, 230°C hydrotreater temperature, and 20:1 bottoms to overhead distillation ratio. Process economic feasibility occurs if cheap CO2, H2O and nutrient resources are available, along with solar energy and energy from byproduct combustion, and hydrotreater H2 from product reforming. Copyright © 2014 Elsevier Ltd. All rights reserved.

  3. Mechanical separation process for decladding of LWR fuel elements

    Koch, R.

    1984-10-01

    A comparison of the advantages and disadvantages of known methods of decladding led to cavitation erosion being used as a decladding mechanism. This process attacks not the jacket of the fuel rod but the fuel itself. Cavitation erosion is the consequence of imploding vapour bubbles entailing dynamic stress of a high frequency and high amplitude. The separation effect is due to the different material properties. Ductile materials as a rule are much more resistant to dynamic stress than brittle materials. Systematic experiments at varying pressures, volume flow, nozzle geometries and distances between nozzle and sample led to optimized parameters. There was a conspicuous rise in the relations pressure to depth of erosion and volume flow to depth of erosion. This considered, p=700 bar and d=1.6 mm were found to be useful parameters. The relation of the distance from nozzle to sample and the erosion obtained also has an optimum at s=50 mm. This distance can be shortened in the course of the operation. A great entrance angle combined with a nozzle outlet channel of the length l=1/2 D improves the erosion result considerably. The attack of the cavitating water jet on the jacket of the fuel rod causes a weight loss of [de

  4. Spent Nuclear Fuel Project Configuration Management Plan

    Reilly, M.A.

    1995-01-01

    This document is a rewrite of the draft ''C'' that was agreed to ''in principle'' by SNF Project level 2 managers on EDT 609835, dated March 1995 (not released). The implementation process philosphy was changed in keeping with the ongoing reengineering of the WHC Controlled Manuals to achieve configuration management within the SNF Project

  5. Spent Nuclear Fuel project interface control plan

    Reilly, M.A.

    1995-01-01

    This implementation process philosophy is in keeping with the ongoing reengineering of the WHC Controlled Manuals to achieve interface control within the SNF Project. This plan applies to all SNF Project sub-project to sub-project, and sub-project to exteranl (both on and off the Hanford Site) interfaces

  6. Advanced Fuel Cycle Cost Basis – 2017 Edition

    Dixon, B. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ganda, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Williams, K. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanson, J. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-09-29

    This report, commissioned by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the DOE Nuclear Technology Research and Development (NTRD) Program (previously the Fuel Cycle Research and Development (FCRD) and the Advanced Fuel Cycle Initiative (AFCI)). The report describes the NTRD cost basis development process, reference information on NTRD cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for numerous fuel cycle cost modules (modules A-O) as well as cost modules for a number of reactor types (R modules). The fuel cycle cost modules were developed in the areas of natural uranium mining and milling, thorium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, managed decay storage, recycled product storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste. Since its inception, this report has been periodically updated. The last such internal document was published in August 2015 while the last external edition was published in December of 2009 as INL/EXT-07-12107 and is available on the Web at URL: www.inl.gov/technicalpublications/Documents/4536700.pdf. This current report (Sept 2017) is planned to be reviewed for external release, at which time it will replace the 2009 report as an external publication. This information is used in the ongoing evaluation of nuclear fuel cycles by the NE NTRD program.

  7. Advanced hybrid process with solvent extraction and pyro-chemical process of spent fuel reprocessing for LWR to FBR

    Fujita, Reiko; Mizuguchi, Koji; Fuse, Kouki; Saso, Michitaka; Utsunomiya, Kazuhiro; Arie, Kazuo

    2008-01-01

    Toshiba has been proposing a new fuel cycle concept of a transition from LWR to FBR. The new fuel cycle concept has better economical process of the LWR spent fuel reprocessing than the present Purex Process and the proliferation resistance for FBR cycle of plutonium with minor actinides after 2040. Toshiba has been developing a new Advanced Hybrid Process with Solvent Extraction and Pyrochemical process of spent fuel reprocessing for LWR to FBR. The Advanced Hybrid Process combines the solvent extraction process of the LWR spent fuel in nitric acid with the recovery of high pure uranium for LWR fuel and the pyro-chemical process in molten salts of impure plutonium recovery with minor actinides for metallic FBR fuel, which is the FBR spent fuel recycle system after FBR age based on the electrorefining process in molten salts since 1988. The new Advanced Hybrid Process enables the decrease of the high-level waste and the secondary waste from the spent fuel reprocessing plants. The R and D costs in the new Advanced Hybrid Process might be reduced because of the mutual Pyro-chemical process in molten salts. This paper describes the new fuel cycle concept of a transition from LWR to FBR and the feasibility of the new Advanced Hybrid Process by fundamental experiments. (author)

  8. Standard guide for drying behavior of spent nuclear fuel

    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...

  9. A novel Snf2 protein maintains trans-generational regulatory states established by paramutation in maize.

    Christopher J Hale

    2007-10-01

    Full Text Available Paramutations represent heritable epigenetic alterations that cause departures from Mendelian inheritance. While the mechanism responsible is largely unknown, recent results in both mouse and maize suggest paramutations are correlated with RNA molecules capable of affecting changes in gene expression patterns. In maize, multiple required to maintain repression (rmr loci stabilize these paramutant states. Here we show rmr1 encodes a novel Snf2 protein that affects both small RNA accumulation and cytosine methylation of a proximal transposon fragment at the Pl1-Rhoades allele. However, these cytosine methylation differences do not define the various epigenetic states associated with paramutations. Pedigree analyses also show RMR1 does not mediate the allelic interactions that typically establish paramutations. Strikingly, our mutant analyses show that Pl1-Rhoades RNA transcript levels are altered independently of transcription rates, implicating a post-transcriptional level of RMR1 action. These results suggest the RNA component of maize paramutation maintains small heterochromatic-like domains that can affect, via the activity of a Snf2 protein, the stability of nascent transcripts from adjacent genes by way of a cotranscriptional repression process. These findings highlight a mechanism by which alleles of endogenous loci can acquire novel expression patterns that are meiotically transmissible.

  10. Pyrolytic electrochemical process for the reprocessing of irradiated nuclear fuels

    Brambilla, G.; Sartorelli, A.

    1980-01-01

    The reprocessing is aimed at synthetic UO 2 -PuO 2 mixed oxides, UC-PuC mixed carbides and at oxides and carbides of U, Pu and Th from fast nuclear reactors. The nuclear fuel is dissolved in a salt melting bath. The conversion of the Pu(SO 4 ) 2 is done thermally and that of UO 2 is done electrolytically. The molten salts are returned to the input of the process and the fission products and the molten salts are conditioned. (DG) [de

  11. Safeguards System for the Advanced Spent Fuel Conditioning Process Facility

    Kim, Ho-dong; Lee, T.H.; Yoon, J.S.; Park, S.W; Lee, S.Y.; Li, T.K.; Menlove, H.; Miller, M.C.; Tolba, A.; Zarucki, R.; Shawky, S.; Kamya, S.

    2007-01-01

    The advanced spent fuel conditioning process (ACP) which is a part of a pyro-processing has been under development at Korean Atomic Energy Research Institute (KAERI) since 1997 to tackle the problem of an accumulation of spent fuel. The concept is to convert spent oxide fuel into a metallic form in a high temperature molten salt in order to reduce the heat energy, volume, and radioactivity of a spent fuel. Since the inactive tests of the ACP have been successfully implemented to confirm the validity of the electrolytic reduction technology, a lab-scale hot test will be undertaken in a couple of years to validate the concept. For this purpose, the KAERI has built the ACP Facility (ACPF) at the basement of the Irradiated Material Examination Facility (IMEF) of KAERI, which already has a reserved hot-cell area. Through the bilateral arrangement between US Department of Energy (DOE) and Korean Ministry of Science and Technology (MOST) for safeguards R and D, the KAERI has developed elements of safeguards system for the ACPF in cooperation with the Los Alamos National Laboratory (LANL). The reference safeguards design conditions and equipment were established for the ACPF. The ACPF safeguards system has many unique design specifications because of the particular characteristics of the pyro-process materials and the restrictions during a facility operation. For the material accounting system, a set of remote operation and maintenance concepts has been introduced for a non-destructive assay (NDA) system. The IAEA has proposed a safeguards approach to the ACPF for the different operational phases. Safeguards measures at the ACPF will be implemented during all operational phases which include a 'Cold Test', a 'Hot Test' and at the end of a 'Hot test'. Optimization of the IAEA's inspection efforts was addressed by designing an effective safeguards approach that relies on, inter alia, remote monitoring using cameras, installed NDA instrumentation, gate monitors and seals

  12. Microbial fuel cell treatment of ethanol fermentation process water

    Borole, Abhijeet P [Knoxville, TN

    2012-06-05

    The present invention relates to a method for removing inhibitor compounds from a cellulosic biomass-to-ethanol process which includes a pretreatment step of raw cellulosic biomass material and the production of fermentation process water after production and removal of ethanol from a fermentation step, the method comprising contacting said fermentation process water with an anode of a microbial fuel cell, said anode containing microbes thereon which oxidatively degrade one or more of said inhibitor compounds while producing electrical energy or hydrogen from said oxidative degradation, and wherein said anode is in electrical communication with a cathode, and a porous material (such as a porous or cation-permeable membrane) separates said anode and cathode.

  13. Zeolites as Catalysts for Fuels Refining after Indirect Liquefaction Processes

    Arno de Klerk

    2018-01-01

    Full Text Available The use of zeolite catalysts for the refining of products from methanol synthesis and Fisher–Tropsch synthesis was reviewed. The focus was on fuels refining processes and differences in the application to indirect liquefaction products was compared to petroleum, which is often a case of managing different molecules. Processes covered were skeletal isomerisation of n-butenes, hydroisomerisation of n-butane, aliphatic alkylation, alkene oligomerisation, methanol to hydrocarbons, ethanol and heavier alcohols to hydrocarbons, carbonyls to hydrocarbons, etherification of alkenes with alcohols, light naphtha hydroisomerisation, catalytic naphtha reforming, hydroisomerisation of distillate, hydrocracking and fluid catalytic cracking. The zeolite types that are already industrially used were pointed out, as well as zeolite types that have future promise for specific conversion processes.

  14. NEW MATERIAL NEEDS FOR HYDROCARBON FUEL PROCESSING: Generating Hydrogen for the PEM Fuel Cell

    Farrauto, R.; Hwang, S.; Shore, L.; Ruettinger, W.; Lampert, J.; Giroux, T.; Liu, Y.; Ilinich, O.

    2003-08-01

    The hydrogen economy is fast approaching as petroleum reserves are rapidly consumed. The fuel cell promises to deliver clean and efficient power by combining hydrogen and oxygen in a simple electrochemical device that directly converts chemical energy to electrical energy. Hydrogen, the most plentiful element available, can be extracted from water by electrolysis. One can imagine capturing energy from the sun and wind and/or from the depths of the earth to provide the necessary power for electrolysis. Alternative energy sources such as these are the promise for the future, but for now they are not feasible for power needs across the globe. A transitional solution is required to convert certain hydrocarbon fuels to hydrogen. These fuels must be available through existing infrastructures such as the natural gas pipeline. The present review discusses the catalyst and adsorbent technologies under development for the extraction of hydrogen from natural gas to meet the requirements for the proton exchange membrane (PEM) fuel cell. The primary market is for residential applications, where pipeline natural gas will be the source of H2 used to power the home. Other applications including the reforming of methanol for portable power applications such as laptop computers, cellular phones, and personnel digital equipment are also discussed. Processing natural gas containing sulfur requires many materials, for example, adsorbents for desulfurization, and heterogeneous catalysts for reforming (either autothermal or steam reforming) water gas shift, preferential oxidation of CO, and anode tail gas combustion. All these technologies are discussed for natural gas and to a limited extent for reforming methanol.

  15. AECL's progress in developing the DUPIC fuel fabrication process

    Sullivan, J.D.; Cox, D.S.

    1995-01-01

    Spent Pressurized Water Reactor (PWR) fuel can be used directly in CANDU reactors without the need for wet chemical reprocessing or reenrichment. Considerable experimental progress has been made in verifying the practicality of this fuel cycle, including hot-cell experiments using spent PWR fuels and out-cell trials using surrogate fuels. This paper describes the current status of these experiments. (author)

  16. Geochemical Interactions in failed Co-Disposal Waste Packages for N Reactor and Ft. St. Vrain Spent Fuel and the Melt and Dilute Waste Form

    Arthur, S.E.; McNeish, J.

    2002-01-01

    The objective of this scientific analysis is to calculate the long-term geochemical behavior in a failed co-disposal waste package (WP) containing U. S. Department of Energy (DOE) spent nuclear fuel (SNF) and high level waste (HLW) glass. This analysis was prepared according to a Technical Work Plan (BSC 2002). Specifically the scope of these calculations is to determine: (1) The geochemical characteristics of the fluids inside the WP after breach, including the corrosion/dissolution of the initial WP configuration; (2) The transport of radionuclides of concern to performance assessment out of the degraded WP by infiltrating water; and (3) The range of parameter variation for additional laboratory and numerical evaluations. This analysis is limited to three SNF groups, uranium (U)/thorium (Th) carbide SNF (Group 5), U metal SNF (Group 7), and aluminum(Al)-based fuels (Group 9). Group 5 is represented by Ft. St. Vrain (FSV) U/Th carbide SNF, Group 7 is represented by N-Reactor U metal SNF, and Group 9 is represented by the Melt and Dilute (MandD) waste form developed from Al-based SNF. The DOE (2001a, Appendix A) describes all of these fuels. Table 1 shows the groups of DOE SNF, the representative SNF for each group, and the metric tons of heavy metal (MTHM) of SNF in each group

  17. Closing the gap between spent fuel storage and final disposal in a multinational management system

    Bredell, P.J.

    1999-01-01

    In this paper, a multinational spent fuel management concept is proposed. The management concept is based on a service agreement between countries, which intend participating in a common spent fuel (SNF) management venture. Accordingly, one of the participants in this venture would act as the hosting country, while the others fulfil the role of customer countries. The hosting country would agree to accept SNF from customer countries under specific conditions, as required by the service agreement. The service agreement should cover a sufficient number of options that customers can use, such as storage, reprocessing or disposal. The service offering should be flexible enough to accommodate diverse customer requirements. Typically, the first step in the multinational management process is the storage of the SNF delivered to the hosting country. The final step being the disposal of the material in a deep geologic repository. This paper explores the ways and means of closing the gap between the first and last steps in the management process. (author)

  18. Radionuclide migration at sites of temporary storage of SNF and RW in North-West Russia - Contribution to regulatory development

    Sneve, M.K.; Shandala, N.K.; Orlova, E.I.; Titov, A.V.; Kochetkov, O.A.; Smith, G.M.; Barraclough, I.M.

    2007-01-01

    Two technical bases of the Northern Fleet were created in the Russian northwest in the 1960s at Andreeva in the Kola Bay and Gremikha village on the coast of the Barents Sea. They maintained nuclear submarines, performing receipt and storage of radioactive waste and spent nuclear fuel. No further stored material was received after 1985. These technical bases have since been re-categorised as sites of temporary storage. It is necessary to note that, during the storage of RW and SNF, certain conditions arose which resulted in failure of the storage barrier system, resulting in release of radionuclides. Remediation activities at the site focus on reduction of major risks associated with most hazardous radioactive source terms. In addition, the long term management of the sites includes consideration of how to remediate contaminated areas, not only because they affect continuing work at the site, but also because this work will influence final radiological status of the sites. The optimum approach to remediation will be affected by how quickly radionuclides could move, both during the remediation works and, so far as any residual activity is concerned, after the works are completed. Present investigations reported here are directed to determination of sorption-desorption parameters of radionuclides in the studied areas, which will affect their underground migration, with the purpose of accounting for regional peculiarities in optimization process of the STSs remediation. The work is being carried out by the TSO State Research Centre - Institute of Biophysics, of Russian Federation, with assistance from western experts. The work forms part of a regulatory collaboration programme on-going between the Norwegian Radiation Protection Authority and the Federal Medical-Biological Agency which is designed to support the development of norms and standards to be applied in the remediation of these sites of temporary storage. (author)

  19. Spent nuclear fuel project multi-year work plan WBS {number_sign}1.4.1

    Wells, J.L.

    1997-03-01

    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.

  20. Development of advanced spent fuel management process. The fabrication and oxidation behavior of simulated metallized spent fuel

    Ro, Seung Gy; Shin, Y.J.; You, G.S.; Joo, J.S.; Min, D.K.; Chun, Y.B.; Lee, E.P.; Seo, H.S.; Ahn, S.B

    1999-03-01

    The simulated metallized spent fuel ingots were fabricated and evaluated the oxidation rates and the activation energies under several temperature conditions to develop an advanced spent fuel management process. It was also checked the alloying characteristics of the some elements with metal uranium. (Author). 3 refs., 1 tab., 36 figs.

  1. Rework of process effluents from the fabrication of HTR fuel

    Lasberg, Ingo; Braehler, Georg [NUKEM Technologies GmbH (Germany); Boyes, David [Pebble Bed Modular Reactor (Pty) Ltd., Centurion (South Africa)

    2008-07-01

    HTR fuel facilities require the application of several liquid chemicals and accordingly they produce significant amounts of Uranium contaminated/potentially contaminated effluents. The main effluents are (amounts for a 3 t Uranium/a plant): aqueous solutions including tetrahydrofurfuryl alcohol THFA, ammonium hydroxide NH4OH, and ammonium nitrate NH4NO3 (180 m{sup 3}/a), isopropanol IPA/water mixtures (130 m{sup 3}/a); Non-Process Water NPW (300 m{sup 3}/a); methanol (7m{sup 3}/a); additionally off-gas streams, containing ammonia (9 t/a) have to be treated. In an industrial scale facility all such effluents/gases need to be processed for recycling, decontamination prior to release to the environment (as waste or as valuable material). Thermal decomposition is applied to dispose of burnable residues.

  2. Process engineering of ceramic composite coatings for fuel cell systems

    Li, G.; Kim, H.; Chen, M.; Yang, Q.; Troczynski, T. [British Columbia Univ., Vancouver, BC (Canada). Dept. of Metals and Materials Engineering

    2003-07-01

    Researchers at UBCeram at the Department of Metals and Materials Engineering at the University of British Columbia have developed a technology to chemically bond composite sol-gel (CB-CSG) coating onto metallic surfaces of complex or concave shapes. The process has been optimized for electrically resistive coatings and corrosion-resistant coatings. The CSG is sprayed onto metallic surfaces and is heat-treated at 300 degrees C to partially dehydrate the hydroxides. The CSG film is then chemically bonded through reaction of active alumina with metal phosphates, such as aluminium phosphate. A new chromate-free process is being developed to address the issue of coatings porosity. The electrodeposition technique involves polymer particles mixed with suspended fine alumina particles which are co-deposited by electrophoretic means or by electrocoagulation. The composite e-coatings have excellent mechanical properties and are being considered as a protective coating for various components of fuel cell systems. 9 refs., 7 figs.

  3. Synthesis of Diopside by Solution Combustion Process Using Glycine Fuel

    Sherikar, Baburao N.; Umarji, A. M.

    Nano ceramic Diopside (CaMgSi2O6) powders are synthesized by Solution Combustion Process(SCS) using Calcium nitrate, Magnesium nitrate as oxidizer and glycine as fuel, fumed silica as silica source. Ammonium nitrate (AN) is used as extra oxidizer. Effect of AN on Diopside phase formation is investigated. The adiabatic flame temperatures are calculated theoretically for varying amount of AN according to thermodynamic concept and correlated with the observed flame temperatures. A “Multi channel thermocouple setup connected to computer interfaced Keithley multi voltmeter 2700” is used to monitor the thermal events during the process. An interpretation based on maximum combustion temperature and the amount of gases produced during reaction for various AN compositions has been proposed for the nature of combustion and its correlation with the characteristics of as synthesized powder. These powders are characterized by XRD, SEM showing that the powders are composed of polycrystalline oxides with crystallite size of 58nm to 74nm.

  4. Role of thermal analysis in uranium oxide fuel fabrication process

    Balaji Rao, Y.; Yadav, R.B.

    2006-01-01

    The present paper discusses the application of thermal analysis, particularly, differential thermal analysis (Dta) at various stages of fuel fabrication process. The useful role of Dta in knowing the decomposition pattern and calcination temperature of Adu along with de-nitration temperature is explained. The decomposition pattern depends upon the type of drying process adopted for wet ADU cake (ADU C). Also, the paper highlights the utility of DTA in determining the APS and SSA of UO 2+x and U 3 O 8 powders as an alternate technique. Further, the temperature difference (ΔT max ) between the two exothermic peaks obtained in UO 2+x powder oxidation is related to sintered density of UO 2 pellets. (author)

  5. Rework of process effluents from the fabrication of HTR fuel

    Lasberg, Ingo; Braehler, Georg; Boyes, David

    2008-01-01

    HTR fuel facilities require the application of several liquid chemicals and accordingly they produce significant amounts of Uranium contaminated/potentially contaminated effluents. The main effluents are (amounts for a 3 t Uranium/a plant): aqueous solutions including tetrahydrofurfuryl alcohol THFA, ammonium hydroxide NH4OH, and ammonium nitrate NH4NO3 (180 m 3 /a), isopropanol IPA/water mixtures (130 m 3 /a); Non-Process Water NPW (300 m 3 /a); methanol (7m 3 /a); additionally off-gas streams, containing ammonia (9 t/a) have to be treated. In an industrial scale facility all such effluents/gases need to be processed for recycling, decontamination prior to release to the environment (as waste or as valuable material). Thermal decomposition is applied to dispose of burnable residues.

  6. Manufacturing process for the metal ceramic hybrid fuel cladding tube

    Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong

    2012-01-01

    For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced

  7. Liquid fuels from biomass via a hydrothermal process

    Goudriaan, F.; Peferoen, D.G.R. (Koninklijke Shell, Amsterdam (Netherlands). Lab.)

    1990-01-01

    Preliminary process studies on the conversion of various biomass types into liquid fuels have indicated that HydroThermal Upgrading (HTU) is more attractive than pyrolysis or gasification. In HTU the biomass is treated at temperatures of 300-350{sup 0}C in the presence of liquid water for 5-15 min. A large proportion of the oxygen is removed as carbon dioxide. In a case study a process for the production of 3600 t/d hydrocarbons starting from wood is evaluated. Six HTU units convert wood into ''biocrude'' containing 10 %w oxygen. The biocrude is upgraded by catalytic hydrodeoxygenation in a central facility. The final products are kerosine and gas oil which may be expected to have excellent properties. The manufacturing cost is 400-450 $/t. (author).

  8. Processing biogas to obtain motor fuel - Operational experience

    Seifert, M.

    2008-01-01

    This article takes a look at how raw biogas can be processed in order to remove carbon dioxide and corrosive substances and thus bring it up to natural gas quality. The ecological advantages of using biogas as a fuel are discussed and the situation in Europe and Switzerland is examined. Also, feeding biogas into the normal natural gas mains is discussed and the technologies necessary for the cleaning and preparation of the biogas are described. These include absorption and adsorption processes as well as membrane systems that are used to remove excessive carbon dioxide. The costs involved are discussed on the basis of experience gained in Sweden and Switzerland. Finally, the environmental aspects of methane losses are discussed.

  9. Manufacturing process for the metal ceramic hybrid fuel cladding tube

    Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced.

  10. Process for the production of fuel combined articles for addition in block shaped high temperature fuel elements

    Hrovat, M.; Rachor, L.

    1976-01-01

    There is provided a process for the production of fuel compacts consisting of an isotropic, radiation-resistant graphite matrix of good heat conductivity having embedded therein coated fuel and/or fertile particles for insertion into high temperature fuel elements by providing the coated fuel and/or fertile particles with an overcoat of molding mixture consisting of graphite powder and a thermoplastic resin binder. The particles after the overcoating are provided with hardener and lubricant only on the surface and subsequently are compressed in a die heated to a constant temperature of about 150 0 C, hardened and discharged therefrom as finished compacts

  11. CRITICALITY CALCULATION FOR THE MOST REACTIVE DEGRADED CONFIGURATIONS OF THE FFTF SNF CODISPOSAL WP CONTAINING AN INTACT IDENT-69 CONTAINER

    D.R. Moscalu

    2002-01-01

    The objective of this calculation is to perform additional degraded mode criticality evaluations of the Department of Energy's (DOE) Fast Flux Test Facility (FFTF) Spent Nuclear Fuel (SNF) codisposed in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP). The scope of this calculation is limited to the most reactive degraded configurations of the codisposal WP with an almost intact Ident-69 container (breached and flooded but otherwise non-degraded) containing intact FFTF SNF pins. The configurations have been identified in a previous analysis (CRWMS M andO 1999a) and the present evaluations include additional relevant information that was left out of the original calculations. The additional information describes the exact distribution of fissile material in each container (DOE 2002a). The effects of the changes that have been included in the baseline design of the codisposal WP (CRWMS M andO 2000) are also investigated. The calculation determines the effective neutron multiplication factor (k eff ) for selected degraded mode internal configurations of the codisposal waste package. These calculations will support the demonstration of the technical viability of the design solution adopted for disposing of MOX (FFTF) spent nuclear fuel in the potential repository. This calculation is subject to the Quality Assurance Requirements and Description (QARD) (DOE 2002b) per the activity evaluation under work package number P6212310M2 in the technical work plan TWP-MGR-MD-0000101 (BSC 2002)

  12. Report of Post Irradiation Examination for Dry Process Fuel

    Par, Jang Jin; Jung, I. H.; Kang, K. H.; Moon, J. S.; Lee, C. R.; Ryu, H. J.; Song, K. C.; Yang, M. S.; Yoo, B. O.; Jung, Y. H.; Choo, Y. S.

    2006-08-01

    The spent PWR fuel typically contains 0.9 wt.% of fissile uranium and 0.6 wt.% of fissile plutonium, which exceeds the natural uranium fissile content of 0.711 wt.%. The neutron economy of a CANDU reactor is sufficient to utilize the DUPIC fuel, even though the neutron-absorbing fission products contained in the spent PWR fuel were remained in the DUPIC fuel. The DUPIC fuel cycle offers advantages to the countries operating both the PWR and CANDU reactors, such as saving the natural uranium, reducing the spent fuel in both PWR and CANDU, and acquiring the extra energy by reuse of the PWR spent fuel. This report contains the results of post-irradiation examination of the DUPIC fuel irradiated four times at HANARO from May 2000 to August 2006 present except the first irradiation test of simulated DUPIC fuel at HANARO on August 1999

  13. The remote methods for radwaste and SNF control

    Ivanov, O; Stepanov, V; Danilovich, A; Potapov, V

    2017-01-01

    With the examples of developments carried out in the Kurchatov Institute and by the world leaders in the field the presentation considers the devices and methods to obtain remotely information on the distribution of radioactivity in radwaste and SNF. It describes the different types of light portable gamma cameras. The application of scanning spectrometric systems is considers also. The methods of recording UV radiation for detection of alpha contamination with the luminescence of air are presented. We discuss the scope and tasks that can be solved using remote and non-destructive methods. (paper)

  14. Radioactive Release from Aluminum-Based Spent Nuclear Fuel in Basin Storage

    Sindelar, R.L.

    1999-10-21

    The report provides an evaluation of: (1) the release rate of radionuclides through minor cladding penetrations (breaches) on aluminum-based spent nuclear fuel (AL SNF), and (2) the consequences of direct storage of breached AL SNF relative to the authorization basis for SRS basin operation.

  15. Characterization program management plan for Hanford K basin spent nuclear fuel

    TRIMBLE, D.J.

    1999-01-01

    The program management plan for characterization of the K Basin spent nuclear fuel was revised to incorporate corrective actions in response to SNF Project QA surveillance 1K-FY-99-060. This revision of the SNF Characterization PMP replaces Duke Eng

  16. Radioactive Release from Aluminum-Based Spent Nuclear Fuel in Basin Storage

    Sindelar, R.L.

    1999-01-01

    The report provides an evaluation of: (1) the release rate of radionuclides through minor cladding penetrations (breaches) on aluminum-based spent nuclear fuel (AL SNF), and (2) the consequences of direct storage of breached AL SNF relative to the authorization basis for SRS basin operation

  17. Subunits of the Snf1 kinase heterotrimer show interdependence for association and activity.

    Elbing, Karin; Rubenstein, Eric M; McCartney, Rhonda R; Schmidt, Martin C

    2006-09-08

    The Snf1 kinase and its mammalian orthologue, the AMP-activated protein kinase (AMPK), function as heterotrimers composed of a catalytic alpha-subunit and two non-catalytic subunits, beta and gamma. The beta-subunit is thought to hold the complex together and control subcellular localization whereas the gamma-subunit plays a regulatory role by binding to and blocking the function of an auto-inhibitory domain (AID) present in the alpha-subunit. In addition, catalytic activity requires phosphorylation by a distinct upstream kinase. In yeast, any one of three Snf1-activating kinases, Sak1, Tos3, or Elm1, can fulfill this role. We have previously shown that Sak1 is the only Snf1-activating kinase that forms a stable complex with Snf1. Here we show that the formation of the Sak1.Snf1 complex requires the beta- and gamma-subunits in vivo. However, formation of the Sak1.Snf1 complex is not necessary for glucose-regulated phosphorylation of the Snf1 activation loop. Snf1 kinase purified from cells lacking the beta-subunits do not contain any gamma-subunit, indicating that the Snf1 kinase does not form a stable alphagamma dimer in vivo. In vitro kinase assays using purified full-length and truncated Snf1 proteins demonstrate that the kinase domain, which lacks the AID, is significantly more active than the full-length Snf1 protein. Addition of purified beta- and gamma-subunits could stimulate the kinase activity of the full-length alpha-subunit but only when all three subunits were present, suggesting an interdependence of all three subunits for assembly of a functional complex.

  18. Underwater Nuclear Fuel Disassembly and Rod Storage Process and Equipment Description. Volume II

    Viebrock, J.M.

    1981-09-01

    The process, equipment, and the demonstration of the Underwater Nuclear Fuel Disassembly and Rod Storage System are presented. The process was shown to be a viable means of increasing spent fuel pool storage density by taking apart fuel assemblies and storing the fuel rods in a denser fashion than in the original storage racks. The assembly's nonfuel-bearing waste is compacted and containerized. The report documents design criteria and analysis, fabrication, demonstration program results, and proposed enhancements to the system

  19. Fuel decontamination at Ringhals 1 with the new decontamination process IcedecTM

    Fredriksson, E.; Ivars, R.; Rosengren, A.; Granath, G.

    2003-01-01

    The new fuel decontamination technique ICEDEC TM , which has been developed by Westinghouse, is based on abrasion of fuel crud with ice particles. A mixture of ice and water is led continuously through the fuel assembly, which is placed in a specially designed fuel decontamination container connected to a closed loop recirculation system. The ice particles scrape off the loose crud from the fuel surfaces and a mixture of crud and water from the melted ice is then led to a filter unit were the crud is separated from the water. In this paper results of fuel decontamination tests of two-year-old and spent fuel assemblies during spring 2001 at Ringhals 1 are presented. The fuel crud was only released when ice particles passed through the fuel assembly and stopped within ten seconds after the feeding of ice particles had ceased. The activity release from the fuel could thus be performed in a controlled way making the process easy to manage and survey. Activity measurements confirmed that about 50% of the loose crud was removed from the fuel surfaces of the two-year-old assembly. Fuel inspection after the decontamination process showed no influence on the fuel integrity. Furthermore, no enhanced personnel radiation dose was involved with the fuel decontamination compared to normal fuel services. (authors)

  20. Fuel distribution process risk analysis in East Borneo

    Laksmita Raizsa

    2018-01-01

    Full Text Available Fuel distribution is an important aspect of fulfilling the customer’s need. It is risky because it can cause tardiness that can cause fuel scarcity. In the process of distribution, many risks are occurring. House of Risk is a method used for mitigating the risk. It identifies seven risk events and nine risk agents. Matrix occurrence and severity are used for eliminating the minor impact risk. House of Risk 1 is used for determining the Aggregate Risk Potential (ARP. Pareto diagram is applied to prioritize risk that must be mitigated by preventive actions based on ARP. It identifies 4 priority risks, namely A8 (Car trouble, A4 (Human Error, A3 (Error deposit via bank and underpayment, and A6 (traffic accident which should be mitigated. House of Risk 2 makes for mapping between the preventive action and risk agent. It gets the Effectiveness to Difficulty Ratio (ETD for mitigating action. Conducting safety talk routine once every three days with ETD 2088 is the primary preventive actions.

  1. Electrochemical processing of spent nuclear fuels: An overview of oxide reduction in pyroprocessing technology

    Eun-Young Choi

    2015-12-01

    Full Text Available The electrochemical reduction process has been used to reduce spent oxide fuel to a metallic form using pyroprocessing technology for a closed fuel cycle in combination with a metal-fuel fast reactor. In the electrochemical reduction process, oxides fuels are loaded at the cathode basket in molten Li2O–LiCl salt and electrochemically reduced to the metal form. Various approaches based on thermodynamic calculations and experimental studies have been used to understand the electrode reaction and efficiently treat spent fuels. The factors that affect the speed of the electrochemical reduction have been determined to optimize the process and scale-up the electrolysis cell. In addition, demonstrations of the integrated series of processes (electrorefining and salt distillation with the electrochemical reduction have been conducted to realize the oxide fuel cycle. This overview provides insight into the current status of and issues related to the electrochemical processing of spent nuclear fuels.

  2. Process for dehydration of oregano using propane gas as fuel

    Carlos O. Velásquez-Santos

    2014-08-01

    Full Text Available The article describes two important issues, the first is the process to design, implement and validate a mechanical dryer of oregano, using propane gas as fuel, and the second is the cost of the process of dehydrated, taking into account the cost of electric energy consumption by the fan and the cost of propane gas consumption by the heat exchanger. To achieve this, it was necessary review the state of the art and the study of the raw material (oregano, were established as premises of design the necessary technical specifications and the variables involved in the process, using conceptual methods and simulation to ensure that it complies with the ISO standard 7925:1999, which defines the requirements for the marketing of dried oregano and processed. Emphasis was made on the percentage of moisture that is 10%, the moisture of the product was found by the azeotropic distillation method, subsequently was validated the functionality and efficiency, comparing the results from an experimental design, then it was obtained the drying curve of oregano with the prototype of drying and it was checked if it meets ISO 7925:1999 standard and the NTC 4423 standard in order to obtain a final product dehydrated with the percentage of humidity appropriate.

  3. Mass Flow Data Comparison for Comprehensive Fuel Cycle Options

    Kim, T.K.; Taiwo, T.A.; Wigeland, R.A.; Dixon, B.W.; Gehin, J.C.; Todosow, M.

    2015-01-01

    One of the key objectives stated in the United States Department of Energy, Nuclear Energy R and D road-map is the development of sustainable nuclear fuel cycles that improve natural resource utilisation and provide adequate capability and capacity to manage wastes produced by the fuel cycle. In order to inform this objective, an evaluation and screening of nuclear fuel cycle options has been conducted. As part of that effort, the entire fuel cycle options space was represented by 40 Evaluation Groups (EGs), and mass flow information for each of the EGs was provided by using an Analysis Example (AE). In this paper, the mass flow data of the 40 AEs are compared to inform on trends in the natural resource utilisation and nuclear waste generation. For the AEs that need enriched uranium support, the natural uranium required is high and the natural resource utilisation is generally lower than 2% regardless of the fuel cycle strategy (i.e., once-through, limited recycle, or continuous recycle). However, the utilisation could be improved by avoiding enriched uranium fuel support. The natural resource utilisation increases to more than 80% by recycling the nuclear fuel continuously without enriched uranium support. The combined mass of spent nuclear fuel (SNF) and high-level waste (HLW), i.e., SNF+HLW mass, is lower by using a continuous recycle option compared to a once-through fuel cycle option, because SNF mass is converted to mass of recycled products and only fission products and other process losses need to be disposed. The combined disposed mass of depleted uranium (DU), recovered uranium (RU) and thorium (RTh), i.e. DU+RU+RTh mass, has a similar trend to the uranium utilisation. For the AEs that need enriched uranium fuel, the DU and RU are the major fraction by mass of the DU+RU+RTh, which are two orders of magnitude higher in mass compared to those for the AEs that do not need enriched uranium fuel. (authors)

  4. Emission of pollutants from the combustion of composite fuels by metallurgical processes

    J. Łabaj

    2015-10-01

    Full Text Available This paper presents the results of the study on emission characteristics of pollutants resulting from combustion process of composite alternative fuels for use in the processes of pyrometallurgy of copper as an alternative fuel to currently used coke breeze. These fuels are mainly based on waste carrier of “C” element, and the composition of the fuel is modelled in order to obtain the appropriate energy and emission parameters as well as strength parameters. These studies confirmed the possibility of using composite fuels as an alternative reducing agent as well as an energy carrier in the processes of pyrometallurgy of copper.

  5. Proposed pyrometallurgical process for rapid recycle of discharged fuel materials from the integral fast reactor

    Burris, L.; Steindler, M.; Miller, W.

    1984-01-01

    The pool-type Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory includes on-site recycle of discharged core and blanket fuel materials. The process and fabrication steps will be demonstrated in the EBR-II Fuel Cycle Facility with IFR fuel irradiated in EBR-II and the Fast Flux Test Facility. The proposed process consists of two major steps: a halide slagging step and an electrorefining step. The fuel is maintained in the metallic form to yield directly a metal product sufficiently decontaminated to allow recycle to the reactor as new fuel. The process is further described and available information to support its feasibility is presented

  6. A proposed pyrometallurgical process for rapid recycle of discharged fuel materials from the Integral Fast Reactor

    Burris, L.; Steindler, M.; Miller, W.

    1984-01-01

    The Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory includes on-site recycle of discharged core and blanket fuel materials. The process and fabrication steps will be demonstrated in the EBR-II Fuel Cycle Facility with IFR fuel irradiated in EBR-II and the Fast Flux Test Facility. The proposed process consists of two major steps -- a halide slagging step and an electrorefining step. The fuel is maintained in the metallic form to yield directly a metal product sufficiently decontaminated to allow recycle to the reactor as new fuel. The process is further described and available information to support its feasibility is presented

  7. Spent nuclear fuel project recommended reaction rate constants for corrosion of N-Reactor fuel

    Cooper, T.D.; Pajunen, A.L.

    1998-01-01

    The US Department of Energy (DOE) established the Spent Nuclear Fuel Project (SNF Project) to address safety and environmental concerns associated with deteriorating spent nuclear fuel presently stored in the Hanford Site's K Basins. The SNF Project has been tasked by the DOE with moving the spent N-Reactor fuel from wet storage to contained dry storage in order to reduce operating costs and environmental hazards. The chemical reactivity of the fuel must be understood at each process step and during long-term dry storage. Normally, the first step would be to measure the N-fuel reactivity before attempting thermal-hydraulic transfer calculations; however, because of the accelerated project schedule, the initial modeling was performed using literature values for uranium reactivity. These literature values were typically found for unirradiated, uncorroded metal. It was fully recognized from the beginning that irradiation and corrosion effects could cause N-fuel to exhibit quite different reactivities than those commonly found in the literature. Even for unirradiated, uncorroded uranium metal, many independent variables affect uranium metal reactivity resulting in a wide scatter of data. Despite this wide reactivity range, it is necessary to choose a defensible model and estimate the reactivity range of the N-fuel until actual reactivity can be established by characterization activities. McGillivray, Ritchie, and Condon developed data and/or models that apply for certain samples over limited temperature ranges and/or reaction conditions (McGillivray 1994, Ritchie 1981 and 1986, and Condon 1983). These models are based upon small data sets and have relatively large correlation coefficients

  8. Supercritical Fluids Processing of Biomass to Chemicals and Fuels

    Olson, Norman K. [Iowa State Univ., Ames, IA (United States)

    2011-09-28

    The main objective of this project is to develop and/or enhance cost-effective methodologies for converting biomass into a wide variety of chemicals, fuels, and products using supercritical fluids. Supercritical fluids will be used both to perform reactions of biomass to chemicals and products as well as to perform extractions/separations of bio-based chemicals from non-homogeneous mixtures. This work supports the Biomass Program’s Thermochemical Platform Goals. Supercritical fluids are a thermochemical approach to processing biomass that, while aligned with the Biomass Program’s interests in gasification and pyrolysis, offer the potential for more precise and controllable reactions. Indeed, the literature with respect to the use of water as a supercritical fluid frequently refers to “supercritical water gasification” or “supercritical water pyrolysis.”

  9. Castor and Pollux - shielded cells for studying fuel treatment processes

    Faudot, G.; Bathellier, A.

    1969-01-01

    CASTOR and POLLUX, two alpha, beta, gamma cells are described in the present paper. They are located in the CEN at Fontenay-aux-Roses (France). They are designed for improvement studies of the various aqueous separation processes used in irradiated fuels reprocessing plants. Located in the same air-tight steel encasement, they arc inter-connected by a pneumatic transfer. These two cells have a similar in-line conception and they include: a gamma shielding in lead of 10 cm of thickness; an inner air-tight box, made with stainless steel and plexiglas, is maintained in lowering in comparison to room pressure. Eleven Hobson model seven master-slave manipulators allow inner manipulations. Then the inner equipment is described briefly. (author) [fr

  10. Environmental assessment for radioisotope heat source fuel processing and fabrication

    1991-07-01

    DOE has prepared an Environmental Assessment (EA) for radioisotope heat source fuel processing and fabrication involving existing facilities at the Savannah River Site (SRS) near Aiken, South Carolina and the Los Alamos National Laboratory (LANL) near Los Alamos, New Mexico. The proposed action is needed to provide Radioisotope Thermoelectric Generators (RTG) to support the National Aeronautics and Space Administration's (NASA) CRAF and Cassini Missions. Based on the analysis in the EA, DOE has determined that the proposed action does not constitute a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, an Environmental Impact Statement is not required. 30 refs., 5 figs

  11. Sp1 and CREB regulate basal transcription of the human SNF2L gene

    Xia Yu; Jiang Baichun; Zou Yongxin; Gao Guimin; Shang Linshan; Chen Bingxi; Liu Qiji; Gong Yaoqin

    2008-01-01

    Imitation Switch (ISWI) is a member of the SWI2/SNF2 superfamily of ATP-dependent chromatin remodelers, which are involved in multiple nuclear functions, including transcriptional regulation, replication, and chromatin assembly. Mammalian genomes encode two ISWI orthologs, SNF2H and SNF2L. In order to clarify the molecular mechanisms governing the expression of human SNF2L gene, we functionally examined the transcriptional regulation of human SNF2L promoter. Reporter gene assays demonstrated that the minimal SNF2L promoter was located between positions -152 to -86 relative to the transcription start site. In this region we have identified a cAMP-response element (CRE) located at -99 to -92 and a Sp1-binding site at -145 to -135 that play a critical role in regulating basal activity of human SNF2L gene, which were proven by deletion and mutation of specific binding sites, EMSA, and down-regulating Sp1 and CREB via RNAi. This study provides the first insight into the mechanisms that control basal expression of human SNF2L gene

  12. Management and inspection of integrity of spent fuel from IRT MEPhI research reactor

    Aden, V.G.; Bulkin, S.Y.; Sokolov, A.V.; Bushuev, A.V.; Redkin, A.F.; Portnov, A.A.

    2002-01-01

    The information on wet storage and dry storage of the spent nuclear fuel (SNF) of the IRT MEPhI reactor and experience from SNF shipment for reprocessing are presented. The procedure and a facility for nondestructive inspection of local power density fields and the burnup of fuel assemblies based on studying the γ-activity of some fission products generated in U 235 and procedure for inspection of the fuel element cladding leak tightness are described. (author)

  13. Laser-based analytical monitoring in nuclear-fuel processing plants

    Hohimer, J.P.

    1978-09-01

    The use of laser-based analytical methods in nuclear-fuel processing plants is considered. The species and locations for accountability, process control, and effluent control measurements in the Coprocessing, Thorex, and reference Purex fuel processing operations are identified and the conventional analytical methods used for these measurements are summarized. The laser analytical methods based upon Raman, absorption, fluorescence, and nonlinear spectroscopy are reviewed and evaluated for their use in fuel processing plants. After a comparison of the capabilities of the laser-based and conventional analytical methods, the promising areas of application of the laser-based methods in fuel processing plants are identified

  14. Crud in the solvent extraction process for spent fuel reprocessing

    Chen Jing

    2004-01-01

    The crud occurred in Purex process is caused by the degradations of extractant and solvent and the existence of insoluble solid particle in the nuclear fuel reprocessing. The crud seriously affects the operation of the extraction column. The present paper reviews the study status on the crud in the Purex process. It is generally accepted that in the Purex process, particularly in the first cycle, the crud occurrence is related to the capillary chemistry phenomena resulting from the deposits of Zr with TBP degradation products HDBP, H 2 MBP, H 3 PO 4 and the insoluble particle RuO 2 and Pd. The occurrence of deposits and the type of crud are tightly related to the molar ratio of HDBP and Zr, and the aqueous pH. In addition, the effect of degradation products from the diluent, such as kerosene, is an unnegligible factor to cause the crud. The crud can be discharged from the extraction equipment with Na 2 CO 3 or oxalic acid. In the study on simulating the crud, the effects of the deposits of Zr with TBP degradation products HDBP, H 2 MBP and H 2 PO 4 , and the insoluble particle RuO 2 and Pd should be considered at the same time. (authors)

  15. Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

    Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V

    2000-07-01

    One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX {yields} MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

  16. Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

    Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V.

    2000-01-01

    One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX → MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

  17. Application of gel microsphere processes to preparation of Sphere-Pac nuclear fuel

    Haas, P.A.; Notz, K.J.; Spence, R.D.

    1978-01-01

    Sphere-Pac fabrication of nuclear fuels using two or more sizes of oxide or carbide spheres is ideally suited to nonproliferation-fuel cycles and remote refabrication. The sizes and compositions of spheres necessary for such fuel cycles have not been commonly prepared; therefore, modifications of sol-gel processes to meet these requirements are being developed and demonstrated

  18. 40 CFR 80.1416 - Petition process for evaluation of new renewable fuels pathways.

    2010-07-01

    ... renewable fuels pathways. 80.1416 Section 80.1416 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY... Petition process for evaluation of new renewable fuels pathways. (a) Pursuant to this section, a party may... fuel pathway has not been evaluated by EPA to determine if it qualifies for a D code pursuant to § 80...

  19. Co-evolution of SNF spliceosomal proteins with their RNA targets in trans-splicing nematodes.

    Strange, Rex Meade; Russelburg, L Peyton; Delaney, Kimberly J

    2016-08-01

    Although the mechanism of pre-mRNA splicing has been well characterized, the evolution of spliceosomal proteins is poorly understood. The U1A/U2B″/SNF family (hereafter referred to as the SNF family) of RNA binding spliceosomal proteins participates in both the U1 and U2 small interacting nuclear ribonucleoproteins (snRNPs). The highly constrained nature of this system has inhibited an analysis of co-evolutionary trends between the proteins and their RNA binding targets. Here we report accelerated sequence evolution in the SNF protein family in Phylum Nematoda, which has allowed an analysis of protein:RNA co-evolution. In a comparison of SNF genes from ecdysozoan species, we found a correlation between trans-splicing species (nematodes) and increased phylogenetic branch lengths of the SNF protein family, with respect to their sister clade Arthropoda. In particular, we found that nematodes (~70-80 % of pre-mRNAs are trans-spliced) have experienced higher rates of SNF sequence evolution than arthropods (predominantly cis-spliced) at both the nucleotide and amino acid levels. Interestingly, this increased evolutionary rate correlates with the reliance on trans-splicing by nematodes, which would alter the role of the SNF family of spliceosomal proteins. We mapped amino acid substitutions to functionally important regions of the SNF protein, specifically to sites that are predicted to disrupt protein:RNA and protein:protein interactions. Finally, we investigated SNF's RNA targets: the U1 and U2 snRNAs. Both are more divergent in nematodes than arthropods, suggesting the RNAs have co-evolved with SNF in order to maintain the necessarily high affinity interaction that has been characterized in other species.

  20. Control system design specification of advanced spent fuel management process units

    Ahn, S. H.; Kim, S. H.; Yoon, J. S

    2003-06-01

    In this study, the design specifications of instrumentation and control system for advanced spent fuel management process units are presented. The advanced spent fuel management process consists of several process units such as slitting device, dry pulverizing/mixing device, metallizer, etc. In this study, the control and operation characteristics of the advanced spent fuel management mockup process devices and the process devices developed in 2001 and 2002 are analysed. Also, a integral processing system of the unit process control signals is proposed, which the operation efficiency is improved. And a redundant PLC control system is constructed which the reliability is improved. A control scheme is proposed for the time delayed systems compensating the control performance degradation caused by time delay. The control system design specification is presented for the advanced spent fuel management process units. This design specifications can be effectively used for the detail design of the advanced spent fuel management process.

  1. Reconstruction of the yeast Snf1 kinase regulatory network reveals its role as a global energy regulator

    Usaite, Renata; Jewett, Michael C; Oliveira, Ana Paula; Yates, John R; Olsson, Lisbeth; Nielsen, Jens

    2009-01-01

    Highly conserved among eukaryotic cells, the AMP-activated kinase (AMPK) is a central regulator of carbon metabolism. To map the complete network of interactions around AMPK in yeast (Snf1) and to evaluate the role of its regulatory subunit Snf4, we measured global mRNA, protein and metabolite levels in wild type, Δsnf1, Δsnf4, and Δsnfsnf4 knockout strains. Using four newly developed computational tools, including novel DOGMA sub-network analysis, we showed the benefits of three-level ome-data integration to uncover the global Snf1 kinase role in yeast. We for the first time identified Snf1's global regulation on gene and protein expression levels, and showed that yeast Snf1 has a far more extensive function in controlling energy metabolism than reported earlier. Additionally, we identified complementary roles of Snf1 and Snf4. Similar to the function of AMPK in humans, our findings showed that Snf1 is a low-energy checkpoint and that yeast can be used more extensively as a model system for studying the molecular mechanisms underlying the global regulation of AMPK in mammals, failure of which leads to metabolic diseases. PMID:19888214

  2. Remote Handling Devices for Disposition of Enriched Uranium Reactor Fuel Using Melt-Dilute Process

    Heckendorn, F.M.

    2001-01-01

    Remote handling equipment is required to achieve the processing of highly radioactive, post reactor, fuel for the melt-dilute process, which will convert high enrichment uranium fuel elements into lower enrichment forms for subsequent disposal. The melt-dilute process combines highly radioactive enriched uranium fuel elements with deleted uranium and aluminum for inductive melting and inductive stirring steps that produce a stable aluminum/uranium ingot of low enrichment

  3. TRISO-coated spent fuel processing using a Grind-Leach head-end

    Spencer, Barry B.; Del Cul, Guillermo D.; Mattus, Catherine H.; Collins, Emory D.

    2005-01-01

    Processing of TRISO-coated HTGR fuels with the grind-leach process requires that the fuel be finely pulverized for efficient and effective acid dissolution of the fuel components. Mechanical size reduction of the fuel is being investigated with jet mill technology as the final milling step. Laboratory experiments were performed with surrogates of crushed fuel compacts that indicate that milling to very small particle sizes is feasible. The size distribution of the milled product is sensitive to the solids feed rate, and the distribution may be bimodal which could support an effective solid-solid separation. (author)

  4. Integrated spent nuclear fuel database system

    Henline, S.P.; Klingler, K.G.; Schierman, B.H.

    1994-01-01

    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

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

    Abbott, D.G.; Abashian, M.S.; Chakraborti, S.; Roberson, K.; Meloin, J.M.

    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

  6. Spent nuclear fuel project technical databook

    Reilly, M.A.

    1998-01-01

    The Spent Nuclear Fuel (SNF) project technical databook provides project-approved summary tables of selected parameters and derived physical quantities, with nominal design and safety basis values. It contains the parameters necessary for a complete documentation basis of the SNF Project technical and safety baseline. The databook is presented in two volumes. Volume 1 presents K Basins SNF related information. Volume 2 (not yet available) will present selected sludge and water information, as it relates to the sludge and water removal projects. The values, within this databook, shall be used as the foundation for analyses, modeling, assumptions, or other input to SNF project safety analyses or design. All analysis and modeling using a parameter available in this databook are required to use and cite the appropriate associated value, and document any changes to those values (i.e., analysis assumptions, equipment conditions, etc). Characterization and analysis efforts are ongoing to validate, or update these values

  7. Spent nuclear fuel project technical databook

    Reilly, M.A.

    1998-07-22

    The Spent Nuclear Fuel (SNF) project technical databook provides project-approved summary tables of selected parameters and derived physical quantities, with nominal design and safety basis values. It contains the parameters necessary for a complete documentation basis of the SNF Project technical and safety baseline. The databook is presented in two volumes. Volume 1 presents K Basins SNF related information. Volume 2 (not yet available) will present selected sludge and water information, as it relates to the sludge and water removal projects. The values, within this databook, shall be used as the foundation for analyses, modeling, assumptions, or other input to SNF project safety analyses or design. All analysis and modeling using a parameter available in this databook are required to use and cite the appropriate associated value, and document any changes to those values (i.e., analysis assumptions, equipment conditions, etc). Characterization and analysis efforts are ongoing to validate, or update these values.

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

    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.

  9. Hanford spent fuel inventory baseline

    Bergsman, K.H.

    1994-01-01

    This document compiles technical data on irradiated fuel stored at the Hanford Site in support of the Hanford SNF Management Environmental Impact Statement. Fuel included is from the Defense Production Reactors (N Reactor and the single-pass reactors; B, C, D, DR, F, H, KE and KW), the Hanford Fast Flux Test Facility Reactor, the Shipping port Pressurized Water Reactor, and small amounts of miscellaneous fuel from several commercial, research, and experimental reactors

  10. The development of spent fuel storage process equipment

    Yoon, Wan Ki; Kim, Ho Dong; Kim, Ki Joon; Kim, Bum Hoe

    1992-02-01

    A nuclear material accounting system were designed to track the transitions of nuclear materials at the spent-fuel technology research facility. It is embedded in a distributed control system real-time structure of the system gives timely on-line accountancy. And performance of AC servo motor with fuzzy logic control and its applicability to spent fuel management were experimentally evaluated. (Author)

  11. Detailed modeling of common rail fuel injection process

    Seykens, X.L.J.; Somers, L.M.T.; Baert, R.S.G.

    2005-01-01

    Modeling of fuel injection equipment is a tool that is used increasingly for explaining or predicting the effect of advanced diesel injection strategies on combustion and emissions. This paper reports on the modeling of the high-pressure part of a research type Heavy Duty Common Rail (CR) fuel

  12. Spent fuel reprocessing system availability definition by process simulation

    Holder, N.; Haldy, B.B.; Jonzen, M.

    1978-05-01

    To examine nuclear fuel reprocessing plant operating parameters such as maintainability, reliability, availability, equipment redundancy, and surge storage requirements and their effect on plant throughput, a computer simulation model of integrated HTGR fuel reprocessing plant operations is being developed at General Atomic Company (GA). The simulation methodology and the status of the computer programming completed on reprocessing head end systems is reported

  13. Fuel Quality/Processing Study. Volume II. Appendix, Task I, literature survey

    O' Hara, J B; Bela, A; Jentz, N E; Klumpe, H W; Kessler, R E; Kotzot, H T; Loran, B I

    1981-04-01

    This activity was begun with the assembly of information from Parsons' files and from contacts in the development and commercial fields. A further more extensive literature search was carried out using the Energy Data Base and the American Petroleum Institute Data Base. These are part of the DOE/RECON system. Approximately 6000 references and abstracts were obtained from the EDB search. These were reviewed and the especially pertinent documents, approximately 300, were acquired in the form of paper copy or microfiche. A Fuel Properties form was developed for listing information pertinent to gas turbine liquid fuel properties specifications. Fuel properties data for liquid fuels from selected synfuel processes, deemed to be successful candidates for near future commercial plants were tabulated on the forms. The processes selected consisted of H-Coal, SRC-II and Exxon Donor Solvent (EDS) coal liquefaction processes plus Paraho and Tosco shale oil processes. Fuel properties analyses for crude and distillate syncrude process products are contained in Section 2. Analyses representing synthetic fuels given refinery treatments, mostly bench scale hydrotreating, are contained in Section 3. Section 4 discusses gas turbine fuel specifications based on petroleum source fuels as developed by the major gas turbine manufacturers. Section 5 presents the on-site gas turbine fuel treatments applicable to petroleum base fuels impurities content in order to prevent adverse contaminant effects. Section 7 relates the environmental aspects of gas turbine fuel usage and combustion performance. It appears that the near future stationary industrial gas turbine fuel market will require that some of the synthetic fuels be refined to the point that they resemble petroleum based fuels.

  14. Fusion-Fission Hybrid for Fissile Fuel Production without Processing

    Fratoni, M; Moir, R W; Kramer, K J; Latkowski, J F; Meier, W R; Powers, J J

    2012-01-02

    Two scenarios are typically envisioned for thorium fuel cycles: 'open' cycles based on irradiation of {sup 232}Th and fission of {sup 233}U in situ without reprocessing or 'closed' cycles based on irradiation of {sup 232}Th followed by reprocessing, and recycling of {sup 233}U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE1) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuel composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in

  15. Thermal imaging of solid oxide fuel cell anode processes

    Pomfret, Michael B.; Kidwell, David A.; Owrutsky, Jeffrey C. [Chemistry Division, U.S. Naval Research Laboratory, Washington, DC 20375 (United States); Steinhurst, Daniel A. [Nova Research Inc., Alexandria, VA 22308 (United States)

    2010-01-01

    A Si-charge-coupled device (CCD), camera-based, near-infrared imaging system is demonstrated on Ni/yttria-stabilized zirconia (YSZ) fragments and the anodes of working solid oxide fuel cells (SOFCs). NiO reduction to Ni by H{sub 2} and carbon deposition lead to the fragment cooling by 5 {+-} 2 C and 16 {+-} 1 C, respectively. When air is flowed over the fragments, the temperature rises 24 {+-} 1 C as carbon and Ni are oxidized. In an operational SOFC, the decrease in temperature with carbon deposition is only 4.0 {+-} 0.1 C as the process is moderated by the presence of oxides and water. Electrochemical oxidation of carbon deposits results in a {delta}T of +2.2 {+-} 0.2 C, demonstrating that electrochemical oxidation is less vigorous than atmospheric oxidation. While the high temperatures of SOFCs are challenging in many respects, they facilitate thermal imaging because their emission overlaps the spectral response of inexpensive Si-CCD cameras. Using Si-CCD cameras has advantages in terms of cost, resolution, and convenience compared to mid-infrared thermal cameras. High spatial ({proportional_to}0.1 mm) and temperature ({proportional_to}0.1 C) resolutions are achieved in this system. This approach provides a convenient and effective analytical technique for investigating the effects of anode chemistry in operating SOFCs. (author)

  16. Thermal imaging of solid oxide fuel cell anode processes

    Pomfret, Michael B.; Steinhurst, Daniel A.; Kidwell, David A.; Owrutsky, Jeffrey C.

    A Si-charge-coupled device (CCD), camera-based, near-infrared imaging system is demonstrated on Ni/yttria-stabilized zirconia (YSZ) fragments and the anodes of working solid oxide fuel cells (SOFCs). NiO reduction to Ni by H 2 and carbon deposition lead to the fragment cooling by 5 ± 2 °C and 16 ± 1 °C, respectively. When air is flowed over the fragments, the temperature rises 24 ± 1 °C as carbon and Ni are oxidized. In an operational SOFC, the decrease in temperature with carbon deposition is only 4.0 ± 0.1 °C as the process is moderated by the presence of oxides and water. Electrochemical oxidation of carbon deposits results in a Δ T of +2.2 ± 0.2 °C, demonstrating that electrochemical oxidation is less vigorous than atmospheric oxidation. While the high temperatures of SOFCs are challenging in many respects, they facilitate thermal imaging because their emission overlaps the spectral response of inexpensive Si-CCD cameras. Using Si-CCD cameras has advantages in terms of cost, resolution, and convenience compared to mid-infrared thermal cameras. High spatial (∼0.1 mm) and temperature (∼0.1 °C) resolutions are achieved in this system. This approach provides a convenient and effective analytical technique for investigating the effects of anode chemistry in operating SOFCs.

  17. Development of recycling processes for clean rejected MOX fuel pellets

    Khot, P.M.; Singh, G.; Shelke, B.K.; Surendra, B.; Yadav, M.K.; Mishra, A.K.; Afzal, Mohd.; Panakkal, J.P.

    2014-01-01

    Highlights: • Dry and wet (MWDD) methods were developed for 100% recycling of CRO (0.4–44% PuO 2 ). • Dry method showed higher productivity and comparable powder/product characteristics. • MWDD batches demonstrated improved powder/product characteristics to that of virgin. • Second/multiple recycling is possible with MWDD with better powder/product characteristics. • MWDD batches prepared by little milling showed better macroscopic homogeneity to that of virgin. - Abstract: The dry and wet recycling processes have been developed for 100% recycling of Clean Reject Oxide (CRO) generated during the fabrication of MOX fuel, as CRO contains significant amount of plutonium. Plutonium being strategic material need to be circumvented from its proliferation issues related to its storage for long period. It was difficult to recycle CRO containing higher Pu content even with multiple oxidation and reduction steps. The mechanical recycling comprising of jaw crushing and sieving has been coupled with thermal pulverization for recycling CRO with higher Pu content in dry recycling technique. In wet recycling, MicroWave Direct Denitration (MWDD) technique has been developed for 100% recycling of CRO. The powder prepared by dry and wet (MWDD) recycling techniques was characterized by XRD and BET techniques and their effects on the pellets were evaluated. (U,21%Pu)O 2 pellets fabricated from virgin powder and MWDD were characterized using optical microscopy and α-autoradiography and the results obtained were compared

  18. Process and device for feeding a cutter of nuclear fuel

    Chaze, Gilbert; Cherel, Guy; Guilloteau, Rene; Tucoulat, Daniel.

    1975-01-01

    A process is described for feeding the cutter of a nuclear fuel tube bundle by means of a mobile magazine. The first operation consists in grabbing the bundle in the storage pool by means of handling gear incorporated in the cutter magazine and introducing the bundle right into the magazine. The second operation is to locate the position of the bundle and to do so a mobile reference plate, of which the relative positions in relation to the magazine and to the cutting section in the cutter are known, is placed near the end of the magazine that will enter into contact with the cutter, the bundle is then moved until its end enters into contact with the reference plate. The next operation consists in adjusting the position of the bundle by moving it out of the magazine by a given length as from the position of the bundle when it was in contact with the reference plate so as to separate the foot end of the bundle by a cut made right next to this foot end. In several successive operations to adjust the cutting positions, the bundle is fed forward in the cutter by a length equal to a tube section. All these operations are carried out without the bundle being separated from the single handling gear built into the magazine [fr

  19. Heterogeneous catalytic process for alcohol fuels from syngas

    Minahan, D.M.; Nagaki, D.A.

    1995-12-31

    This project is focused on the discovery and evaluation of novel heterogeneous catalyst for the production of oxygenated fuel enhancers from synthesis gas. Catalysts have been studied and optimized for the production of methanol and isobutanol mixtures which may be used for the downstream synthesis of MTBE or related oxygenates. Higher alcohols synthesis (HAS) from syngas was studied; the alcohols that are produced in this process may be used for the downstream synthesis of MTBE or related oxygenates. This work has resulted in the discovery of a catalyst system that is highly selective for isobutanol compared with the prior art. The catalysts operate at high temperature (400{degrees}C), and consist of a spinel oxide support (general formula AB{sub 2}O{sub 4}, where A=M{sup 2+} and B = M{sup 3+}), promoted with various other elements. These catalysts operate by what is believed to be an aldol condensation mechanism, giving a product mix of mainly methanol and isobutanol. In this study, the effect of product feed/recycle (methanol, ethanol. n-propanol, isopropanol, carbon dioxide and water) on the performance of 10-DAN-55 (spinel oxide based catalyst) at 400{degrees}C, 1000 psi, GHSV = 12,000 and syngas (H{sub 2}/CO) ratio = 1:2 (alcohol addition) and 1:1 (carbon dioxide and water addition) was studied. The effect of operation at high temperatures and pressures on the performance of an improved catalyst formulation was also examined.

  20. Reduction of environmental pollution from fuel and target manufacturing processes

    Hardt, H.A.

    1976-10-01

    Nuclear fuel and target manufacturing processes in the 300 Area generate potential environmental pollutants. Efforts to eliminate or reduce their harmful effects have been pursued for many years by the Raw Materials and Raw Materials Technology departments with assistance from other groups, primarily the Project and Health Physics departments. This report documents: methods adopted to reduce pollution; cost of these methods; amount of pollution reduction achieved; and other benefits in cost savings or quality improvement for January 1968 through December 1975. Capital funds totaling $915,000 were spent on these programs. Annual cost savings of $65,000 were realized, and incidental but significant improvements in product quality were obtained. In no case was product quality degraded. Reductions in releases of pollutants are summarized for water pollution, air pollution, and land pollution. In addition to these reductions, intangible benefits were realized including reduced corrosion of structures and equipment; improved working conditions for personnel; energy savings, both on and offplant; improved utilization of natural resources; and reduced impact to environment, both on and offplant

  1. Tecnored process - high potential in using different kinds of solid fuels

    José Henrique Noldin Júnior

    2005-12-01

    Full Text Available One important feature of the Brazilian Tecnored ironmaking process is its flexibility to use different types of solid fuels, other than metallurgical coke, as proved in the pilot plant tests by extensively using green petroleum coke, biomasses, high ash cokes, etc. Even if new solid fuels not thus far used are envisaged for a given project, thru the bench scale simulator of the process it is possible to predict the behavior of such solid fuels in the Tecnored furnace and establish the best techno-economical-environmental equation for its use. This paper discusses the key aspects involved in the use of alternative solid fuels in the Tecnored process.

  2. Research on solvent extraction process for reprocessing of Th-U fuel from HTGR

    Bao Borong; Wang Gaodong; Qian Jun

    1992-05-01

    The unique properties of spent fuel from HTGR (high temperature gas cooled reactor) have been analysed. The single solvent extraction process using 30% TBP for separation and purification of Th-U fuel has been studied. In addition, the solvent extraction process for second uranium purification is also investigated to meet different needs of reprocessing and reproduction of Th-U spent fuel from HTGR

  3. The main chemical safety problems in main process of nuclear fuel reprocessing plant

    Song Fengli; Zhao Shangui; Liu Xinhua; Zhang Chunlong; Lu Dan; Liu Yuntao; Yang Xiaowei; Wang Shijun

    2014-01-01

    There are many chemical reactions in the aqueous process of nuclear fuel reprocessing. The reaction conditions and the products are different so that the chemical safety problems are different. In the paper the chemical reactions in the aqueous process of nuclear fuel reprocessing are described and the main chemical safety problems are analyzed. The reference is offered to the design and accident analysis of the nuclear fuel reprocessing plant. (authors)

  4. Design analysis of various transportation package options for BN-350 SNF in terms of nuclear radiation safety in planning for long-terms dry storage

    Aisabekov, A.Z.; Mukenova, S.A.; Tur, E.S.; Tsyngaev, V.M.

    2004-01-01

    Full text: This effort is performed under the BN-350 reactor facility decommissioning project. One of the project tasks - spent nuclear fuel handling - includes the following: fuel packaging into sealed canisters, transportation of the canisters in multi-seat metallo-concrete containers and placement of the containers for a long-term dry storage. The goal of this effort is to computationally validate nuclear and radiation safety of the SNF containers placed for storage both under normal storage conditions and probable accident situations. The basic unit structure and design configurations are presented: assemblies, canisters, transportation containers. The major factors influencing nuclear and radiation safety are presented: fuel burn-up, enrichment, fabrication tolerance, types of fuel assemblies, configuration of assemblies in the canister and canisters in the container, background of assemblies placed in the reactor and cooling pool. Conditions under which the SNF containers will be stored are described and probable accident situations are listed. Proceeding from the conservatism principle, selection of the assemblies posing the greatest nuclear hazard is validated. A neutron effective multiplication factor is calculated for the SNF containers under the normal storage conditions and for the case of emergency. The effective multiplication factor is shown to be within a standard value of 0.95 in any situation. Based on the experimental data on assembly and canister dose rates, canisters posing the highest radiation threat are selected. Activities of sources and gamma-radiation spectral composition are calculated. Distribution of the dose rate outside the containers both under the normal storage conditions and accident situations are calculated. The results obtained are analyzed

  5. Historic American Engineering Record, Idaho National Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex

    Susan Stacy; Julie Braun

    2006-12-01

    Just as automobiles need fuel to operate, so do nuclear reactors. When fossil fuels such as gasoline are burned to power an automobile, they are consumed immediately and nearly completely in the process. When the fuel is gone, energy production stops. Nuclear reactors are incapable of achieving this near complete burn-up because as the fuel (uranium) that powers them is burned through the process of nuclear fission, a variety of other elements are also created and become intimately associated with the uranium. Because they absorb neutrons, which energize the fission process, these accumulating fission products eventually poison the fuel by stopping the production of energy from it. The fission products may also damage the structural integrity of the fuel elements. Even though the uranium fuel is still present, sometimes in significant quantities, it is unburnable and will not power a reactor unless it is separated from the neutron-absorbing fission products by a method called fuel reprocessing. Construction of the Fuel Reprocessing Complex at the Chem Plant started in 1950 with the Bechtel Corporation serving as construction contractor and American Cyanamid Company as operating contractor. Although the Foster Wheeler Corporation assumed responsibility for the detailed working design of the overall plant, scientists at Oak Ridge designed all of the equipment that would be employed in the uranium separations process. After three years of construction activity and extensive testing, the plant was ready to handle its first load of irradiated fuel.

  6. Permission of the alteration in nuclear fuel material processing business in the Kumatori Works of Nuclear Fuel Industries, Ltd. (report)

    1982-01-01

    A report by the Nuclear Safety Commission to the Prime Minister concerning the permission of the alteration in nuclear fuel material processing business in the Kumatori Works, Nuclear Fuel Industries Ltd., was presented. The technical capability of NFI and the safety of the business alteration were confirmed. The items of the safety examination concerning the alteration, made by the Committee on Examination of Nuclear Fuel Safety of NSC, are as follows: the aseismic design and fire prevention measures of the No.2 processing building and the No.3 waste storage building to be newly installed; criticality control concerning the No.2 processing building; the release control of gaseous and liquid wastes; the management of solid wastes; radiation control; the exposure dose of general public. (J.P.N.)

  7. Environmental protection and processes for burning solid fuel. Zashchita okruzhayushchey sredy i protsessy goreniya tverdovo topliva

    1981-01-01

    The anthology's materials are devoted to studying the mechanism of formation of harmful emissions during burning of solid fuels, methods of suppressing processes of formation of nitrogen oxide in boiler devices, and processes of combustion to create power plants with minimal emissions of NO /SUB x/ and SO /SUB x/ and maximum use of organic and mineral components of the fuel.

  8. Spent nuclear fuel retrieval system fuel handling development testing. Final report

    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. Assessment of the dry process fuel sodium-cooled fast reactors

    Roh, Gyu Hong; Choi, Hang Bok

    2004-04-01

    The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed.

  10. Assessment of the dry process fuel sodium-cooled fast reactors

    Roh, Gyu Hong; Choi, Hang Bok

    2004-04-01

    The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed

  11. Studies on the fission products behavior during dissolution process of BWR spent fuel

    Sato, K.; Nakai, E.; Kobayashi, Y.

    1987-01-01

    In order to obtain basic data on fission products behavior in connection with the head end process of fuel reprocessing, especially to obtain better understanding on undissolved residues, small scale dissolution studies were performed by using BWR spent fuel rods which were irradiated as monitoring fuel rods under the monitoring program for LWR fuel assembly performance entitled PROVING TEST ON RELIABILITY OF FUEL ASSEMBLY . The Zircaloy-2 claddings and the fuel pellets were subjected individually to the following studies on 1) release of fission products during dissolution process, 2) characterization of undissolved residues, and 3) analysis of the claddings. This paper presents comprehensive descriptions of the fission products behavior during dissolution process, based on detailed and through PIE conducted by JNFS under the sponsorship of MITI (Ministry of International Trade and Industry)

  12. Development of nitride fuel and pyrochemical process for transmutation of minor actinides

    Arai, Yasuo; Akabori, Mitsuo; Minato, Kazuo; Uno, Masayoshi

    2010-01-01

    Nitride fuel cycle for transmutation of minor actinides has been investigated under the double-strata fuel cycle concept. Mononitride solid solutions containing minor actinides have been prepared and characterised. Thermo-physical properties, such as thermal expansion, heat capacity and thermal diffusivity, have been measured by use of minor actinide nitride and burn-up simulated nitride samples. Irradiation behaviour of nitride fuel has been examined by irradiation tests. Pyrochemical process for treatment of spent nitride fuel has been investigated mainly by electrochemical measurements and nitride formation behaviour in pyrochemical process has been studied for recycled fuel fabrication. Recent results of experimental study on nitride fuel and pyrochemical process are summarised in the paper. (authors)

  13. Advanced spent fuel processing technologies for the United States GNEP programme

    Laidler, J.J.

    2007-01-01

    Spent fuel processing technologies for future advanced nuclear fuel cycles are being developed under the scope of the Global Nuclear Energy Partnership (GNEP). This effort seeks to make available for future deployment a fissile material recycling system that does not involve the separation of pure plutonium from spent fuel. In the nuclear system proposed by the United States under the GNEP initiative, light water reactor spent fuel is treated by means of a solvent extraction process that involves a group extraction of transuranic elements. The recovered transuranics are recycled as fuel material for advanced burner reactors, which can lead in the long term to fast reactors with conversion ratios greater than unity, helping to assure the sustainability of nuclear power systems. Both aqueous and pyrochemical methods are being considered for fast reactor spent fuel processing in the current US development programme. (author)

  14. Handbook on process and chemistry of nuclear fuel reprocessing version 2

    2008-10-01

    Aqueous nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of aqueous reprocessing, because it contributes to establish and develop fuel reprocessing technology and nuclear fuel cycle treating high burn-up UO 2 fuel and spent MOX fuel, and to utilize aqueous reprocessing technology much widely. This handbook is the second edition of the first report, which summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing' from FY 1993 until FY 2000. (author)

  15. Characterization of Hanford K basin spent nuclear fuel and sludge

    Lawrence, L.A.

    1996-01-01

    A characterization plan was prepared to support the Integrated Process Strategy (IPS) for resolution of the safety and environmental concerns associated with the deteriorating Spent Nuclear Fuel (SNF) stored in the Hanford Site K Basins. This plan provides the structure and logic and identifies the information needs to be supported by the characterization activities. The IPS involves removal of the fuel elements from the storage canister and placing them in a container, i.e., Multiple Canister Overpack (MCO) capable of holding multiple tiers of baskets full of fuel. The MCOs will be vacuum dried to remove free water and shipped to the Container Storage Building (CSB) where they will be staged waiting for hot vacuum conditioning. The MCO will be placed in interim storage in the CSB following conditioning and disposition

  16. Rocket Fuel Synthesis by Fisher-Tropsch Process

    National Aeronautics and Space Administration — This study aims to investigate the feasibility of using Fisher Tropsch (FT), a commercial-scale technology that currently produces liquid fuels from syngas (CO &...

  17. Damaged Spent Nuclear Fuel at U.S. DOE Facilities Experience and Lessons Learned

    Brett W. Carlsen; Eric Woolstenhulme; Roger McCormack

    2005-01-01

    From a handling perspective, any spent nuclear fuel (SNF) that has lost its original technical and functional design capabilities with regard to handling and confinement can be considered as damaged. Some SNF was damaged as a result of experimental activities and destructive examinations; incidents during packaging, handling, and transportation; or degradation that has occurred during storage. Some SNF was mechanically destroyed to protect proprietary SNF designs. Examples of damage to the SNF include failed cladding, failed fuel meat, sectioned test specimens, partially reprocessed SNFs, over-heated elements, dismantled assemblies, and assemblies with lifting fixtures removed. In spite of the challenges involved with handling and storage of damaged SNF, the SNF has been safely handled and stored for many years at DOE storage facilities. This report summarizes a variety of challenges encountered at DOE facilities during interim storage and handling operations along with strategies and solutions that are planned or were implemented to ameliorate those challenges. A discussion of proposed paths forward for moving damaged and nondamaged SNF from interim storage to final disposition in the geologic repository is also presented

  18. Fuel cells science and engineering. Materials, processes, systems and technology. Vol. 1

    Stolten, Detlef; Emonts, Bernd (eds.) [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energieforschung (IEF), Brennstoffzellen (IEF-3)

    2012-07-01

    The first volume is divided in four parts and 22 chapters. It is structured as follows: PART I: Technology. Chapter 1: Technical Advancement of Fuel-Cell Research and Development (Dr. Bernd Emonts, Ludger Blum, Thomas Grube, Werner Lehnert, Juergen Mergel, Martin Mueller and Ralf Peters); 2: Single-Chamber Fuel Cells (Teko W. Napporn and Melanie Kuhn); 3: Technology and Applications of Molten Carbonate Fuel Cells (Barbara Bosio, Elisabetta Arato and Paolo Greppi); 4: Alkaline Fuel Cells (Erich Guelzow); 5: Micro Fuel Cells (Ulf Groos and Dietmar Gerteisen); 6: Principles and Technology of Microbial Fuel Cells (Jan B. A. Arends, Joachim Desloover, Sebastia Puig and Willy Verstraete); 7: Micro-Reactors for Fuel Processing (Gunther Kolb); 8: Regenerative Fuel Cells (Martin Mueller). PART II: Materials and Production Processes. Chapter 9: Advances in Solid Oxide Fuel Cell Development between 1995 and 2010 at Forschungszentrum Juelich GmbH, Germany (Vincent Haanappel); 10: Solid Oxide Fuel Cell Electrode Fabrication by Infiltration (Evren Gunen); 11: Sealing Technology for Solid Oxide Fuel Cells (K. Scott Weil); 12: Phosphoric Acid, an Electrolyte for Fuel Cells - Temperature and Composition Dependence of Vapor Pressure and Proton Conductivity (Carsten Korte); 13: Materials and Coatings for Metallic Bipolar Plates in Polymer Electrolyte Membrane Fuel Cells (Heli Wang and John A. Turner); 14: Nanostructured Materials for Fuel Cells (John F. Elter); 15: Catalysis in Low-Temperature Fuel Cells - An Overview (Sabine Schimpf and Michael Bron). PART III: Analytics and Diagnostics. Chapter 16: Impedance Spectroscopy for High-Temperature Fuel Cells (Ellen Ivers-Tiffee, Andre Leonide, Helge Schichlein, Volker Sonn and Andre Weber); 17: Post-Test Characterization of Solid Oxide Fuel-Cell Stacks (Norbert H. Menzler and Peter Batfalsky); 18: In Situ Imaging at Large-Scale Facilities (Christian Toetzke, Ingo Manke and Werner Lehnert); 19: Analytics of Physical Properties of Low

  19. Simulation codes of chemical separation process of spent fuel reprocessing. Tool for process development and safety research

    Asakura, Toshihide; Sato, Makoto; Matsumura, Masakazu; Morita, Yasuji

    2005-01-01

    This paper reviews the succeeding development and utilization of Extraction System Simulation Code for Advanced Reprocessing (ESSCAR). From the viewpoint of development, more tests with spent fuel and calculations should be performed with better understanding of the physico-chemical phenomena in a separation process. From the viewpoint of process safety research on fuel cycle facilities, it is important to know the process behavior of a key substance; being highly reactive but existing only trace amount. (author)

  20. Process and equipment for pressure build-up in nuclear reactor fuel rods

    Heer, W.F.; Carli, E.V. de.

    1976-01-01

    The equipment makes possible the build-up of inert gas pressure in a filled and closed fuel can, i.e. in a complete fuel rod. Handling is simple, it is suitable for mass production and only causes low processing costs. The quality, e.g. the degree of purity of the contents of the rod, remains unchangedin processing. The equipment consists of a vacuum-tight space, into which the equally vacuum tight fuel rod is introduced, and can be fixed so that its position can be reproduced unmistakeably. The vacuum space contains a connection for the inert gases and a laser arrangement. After inserting a fuel rod into the facility, this is evacuated and the fuel can has a hole bored in it by a laser beam. After fast equalisation of pressure, an inert gas at the required pressure is introduced into the chamber and the fuel rod. After the filling process is completed, the fuel can is closed again with the same laser beam. The quality of the seal obtained, i.e the leak-tightness of the fuel can, can be checked after reduction of the inert gas pressure and before taking out the fuel rod, by repeated evacuation of the chamber. Laser light energies between 13,000 and 110,000 Joule/sq cm are sufficient. Optimum results were obtained for a Zircaloy fuel can with about 52,000 Joule/sq cm. (TK) [de