Transportation accident scenarios for commercial spent fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilmot, E L

1981-02-01

A spectrum of high severity, low probability, transportation accident scenarios involving commercial spent fuel is presented together with mechanisms, pathways and quantities of material that might be released from spent fuel to the environment. These scenarios are based on conclusions from a workshop, conducted in May 1980 to discuss transportation accident scenarios, in which a group of experts reviewed and critiqued available literature relating to spent fuel behavior and cask response in accidents.

Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gregson, Michael Warren; Mo, Tin; Sorenson, Ken Bryce

The authors provide a detailed overview of an on-going, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high-energy-density device. The program participants in the United States plus Germany, France and the United Kingdom, part of the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) have strongly supported and coordinated this research program. Sandia National Laboratoriesmore » has the lead role for conducting this research program; test program support is provided by both the US Department of Energy and the US Nuclear Regulatory Commission. The authors provide a summary of the overall, multiphase test design and a description of all explosive containment and aerosol collection test components used. They focus on the recently initiated tests on 'surrogate' spent fuel, unirradiated depleted uranium oxide and forthcoming actual spent fuel tests, and briefly summarize similar results from completed surrogate tests that used non-radioactive, sintered cerium oxide ceramic pellets in test rods.« less

The use of MOX caramel fuel mixed with 241Am, 242mAm and 243Am as burnable absorber actinides for the MTR research reactors.

PubMed

Shaaban, Ismail; Albarhoum, Mohamad

2017-07-01

The MOX (UO 2 &PuO 2 ) caramel fuel mixed with 241 Am, 242m Am and 243 Am as burnable absorber actinides was proposed as a fuel of the MTR-22MW reactor. The MCNP4C code was used to simulate the MTR-22MW reactor and estimate the criticality and the neutronic parameters, and the power peaking factors before and after replacing its original fuel (U 3 O 8 -Al) by the MOX caramel fuel mixed with 241 Am, 242m Am and 243 Am actinides. The obtained results of the criticality, the neutronic parameters, and the power peaking factors for the MOX caramel fuel mixed with 241 Am, 242m Am and 243 Am actinides were compared with the same parameters of the U 3 O 8 -Al original fuel and a maximum difference is -6.18% was found. Additionally, by recycling 2.65% and 2.71% plutonium and 241 Am, 242m Am and 243 Am actinides in the MTR-22MW reactor, the level of 235 U enrichment is reduced from 4.48% to 3% and 2.8%, respectively. This also results in the reduction of the 235 U loading by 32.75% and 37.22% for the 2.65%, the 2.71% plutonium and 241 Am, 242m Am and 243 Am actinides, respectively. Copyright © 2017 Elsevier Ltd. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sarta, Jose A.; Castiblanco, Luis A

With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several calculations and tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safely a program was established at the Instituto de Ciencias Nucleares y Energias Alternativas (INEA). This program included training, acquisition of hardware and software, design and construction of a decay pool, transfer ofmore » the spent HEU fuel elements into the decay pool and his final transport to Savannah River in United States. In this paper are presented data of activities calculated for each relevant radionuclide present in spent MTR-HEU fuel elements of the IAN-R1 Research Reactor and the total activity. The total activity calculated takes in consideration contributions of fission, activation and actinides products. The data obtained were the base for shielding calculations for the decay pool concerning the storage of spent MTR-HEU fuel elements and the respective dosimetric evaluations in the transferring operations of fuel elements into the decay pool.« less

Spent Nuclear Fuel (SNF) Project Execution Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

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.

Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daniel, W. E.; Rudisill, T. S.; O'Rourke, P. E.

2017-01-26

In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgasmore » composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.« less

Spent Nuclear Fuel Transport Reliability Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong; Jiang, Hao

This conference paper was orignated and shorten from the following publisehd PTS documents: 1. Jy-An Wang, Hao Jiang, and Hong Wang, Dynamic Deformation Simulation of Spent Nuclear Fuel Assembly and CIRFT Deformation Sensor Stability Investigation, ORNL/SPR-2015/662, November 2015. 2. Jy-An Wang, Hong Wang, Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications, NUREG/CR-7198, ORNL/TM-2014/214, May 2015. 3. Jy-An Wang, Hong Wang, Hao Jiang, Yong Yan, Bruce Bevard, Spent Nuclear Fuel Vibration Integrity Study 16332, WM2016 Conference, March 6 10, 2016, Phoenix, Arizona.

Reprocessing of research reactor fuel the Dounreay option

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cartwright, P.

1997-08-01

Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storagemore » tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.« less

Method for shearing spent nuclear fuel assemblies

DOEpatents

Weil, Bradley S.; Watson, Clyde D.

1977-01-01

A method is disclosed for shearing spent nuclear fuel assemblies of the type wherein a plurality of long metal tubes packed with ceramic fuel are supported in a spaced apart relationship within an outer metal shell or shroud which provides structural support to the assembly. Spent nuclear fuel assemblies are first compacted in a stepwise manner between specially designed gag-compactors and then sheared into short segments amenable to chemical processing by shear blades contoured to mate with the compacted surface of the fuel assembly.

Rack for storing spent nuclear fuel elements

DOEpatents

Rubinstein, Herbert J.; Clark, Philip M.; Gilcrest, James D.

1978-06-20

A rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed fuel elements. The enclosures are fixed at the lower ends thereof to a base. Pockets are formed between confronting walls of adjacent enclosures for receiving high absorption neutron absorbers, such as Boral, cadmium, borated stainless steel and the like for the closer spacing of spent fuel elements.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2012-06-06

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

Cosmic ray muons for spent nuclear fuel monitoring

NASA Astrophysics Data System (ADS)

Chatzidakis, Stylianos

There is a steady increase in the volume of spent nuclear fuel stored on-site (at reactor) as currently there is no permanent disposal option. No alternative disposal path is available and storage of spent nuclear fuel in dry storage containers is anticipated for the near future. In this dissertation, a capability to monitor spent nuclear fuel stored within dry casks using cosmic ray muons is developed. The motivation stems from the need to investigate whether the stored content agrees with facility declarations to allow proliferation detection and international treaty verification. Cosmic ray muons are charged particles generated naturally in the atmosphere from high energy cosmic rays. Using muons for proliferation detection and international treaty verification of spent nuclear fuel is a novel approach to nuclear security that presents significant advantages. Among others, muons have the ability to penetrate high density materials, are freely available, no radiological sources are required and consequently there is a total absence of any artificial radiological dose. A methodology is developed to demonstrate the applicability of muons for nuclear nonproliferation monitoring of spent nuclear fuel dry casks. Purpose is to use muons to differentiate between spent nuclear fuel dry casks with different amount of loading, not feasible with any other technique. Muon scattering and transmission are used to perform monitoring and imaging of the stored contents of dry casks loaded with spent nuclear fuel. It is shown that one missing fuel assembly can be distinguished from a fully loaded cask with a small overlapping between the scattering distributions with 300,000 muons or more. A Bayesian monitoring algorithm was derived to allow differentiation of a fully loaded dry cask from one with a fuel assembly missing in the order of minutes and negligible error rate. Muon scattering and transmission simulations are used to reconstruct the stored contents of sealed dry casks

Spent Nuclear Fuel Alternative Technology Decision Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 andmore » ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.« less

Spent nuclear fuel assembly inspection using neutron computed tomography

NASA Astrophysics Data System (ADS)

Pope, Chad Lee

The research presented here focuses on spent nuclear fuel assembly inspection using neutron computed tomography. Experimental measurements involving neutron beam transmission through a spent nuclear fuel assembly serve as benchmark measurements for an MCNP simulation model. Comparison of measured results to simulation results shows good agreement. Generation of tomography images from MCNP tally results was accomplished using adapted versions of built in MATLAB algorithms. Multiple fuel assembly models were examined to provide a broad set of conclusions. Tomography images revealing assembly geometric information including the fuel element lattice structure and missing elements can be obtained using high energy neutrons. A projection difference technique was developed which reveals the substitution of unirradiated fuel elements for irradiated fuel elements, using high energy neutrons. More subtle material differences such as altering the burnup of individual elements can be identified with lower energy neutrons provided the scattered neutron contribution to the image is limited. The research results show that neutron computed tomography can be used to inspect spent nuclear fuel assemblies for the purpose of identifying anomalies such as missing elements or substituted elements. The ability to identify anomalies in spent fuel assemblies can be used to deter diversion of material by increasing the risk of early detection as well as improve reprocessing facility operations by confirming the spent fuel configuration is as expected or allowing segregation if anomalies are detected.

A Non-Proliferating Fuel Cycle: No Enrichment, Reprocessing or Accessible Spent Fuel - 12375

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parker, Frank L.

2012-07-01

Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storagemore » sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded emissions

77 FR 76952 - Rescinding Spent Fuel Pool Exclusion Regulations

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-31

... [Docket No. PRM-51-29; NRC-2012-0215] Rescinding Spent Fuel Pool Exclusion Regulations AGENCY: Nuclear... NRC institute a rulemaking to rescind the regulations excluding consideration of spent fuel pool...

An approach to meeting the spent fuel standard

DOE Office of Scientific and Technical Information (OSTI.GOV)

Makhijani, A.

1996-05-01

The idea of the spent fuel standard is that there should be a high surface gamma radiation to prevent theft. For purposes of preventing theft, containers should be massive, and the plutonium should be difficult to extract. This report discusses issues associated with the spent fuel standard.

Numerical Estimation of the Spent Fuel Ratio

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindgren, Eric R.; Durbin, Samuel; Wilke, Jason

Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO 2 ), have been conducted in the interim to more definitively determine the source term from these postulated events. However, the validity of these large- scale results remain in question due to the lack of a defensible spent fuel ratio (SFR), defined as the amount of respirable aerosol generated by an attack on amore » mass of spent fuel compared to that of an otherwise identical surrogate. Previous attempts to define the SFR in the 1980's have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Because of the large uncertainty surrounding the SFR, estimates of releases from security-related events may be unnecessarily conservative. Credible arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and storage of spent nuclear fuel in dry cask systems. In the present work, the shock physics codes CTH and ALE3D were used to simulate spent nuclear fuel (SNF) and DUO 2 targets impacted by a high-velocity jet at an ambient temperature condition. These preliminary results are used to illustrate an approach to estimate the respirable release fraction for each type of material and ultimately, an estimate of the SFR. This page intentionally blank« less

Analysis of Structural MtrC Models Based on Homology with the Crystal Structure of MtrF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwards, Marcus; Fredrickson, Jim K.; Zachara, John M.

2012-12-01

The outer-membrane decahaem cytochrome MtrC is part of the transmembrane MtrCAB complex required for mineral respiration by Shewanella oneidensis. MtrC has significant sequence similarity to the paralogous decahaem cytochrome MtrF, which has been structurally solved through X-ray crystallography. This now allows for homology-based models of MtrC to be generated. The structure of these MtrC homology models contain ten bis-histidine-co-ordinated c-type haems arranged in a staggered cross through a four-domain structure. This model is consistent with current spectroscopic data and shows that the areas around haem 5 and haem 10, at the termini of an octahaem chain, are likely to havemore » functions similar to those of the corresponding haems in MtrF. The electrostatic surfaces around haem 7, close to the β-barrels, are different in MtrF and MtrC, indicating that these haems may have different potentials and interact with substrates differently.« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-13

... radiological impacts of spent nuclear fuel and high-level waste disposal. DATES: Submit comments on the... determination. The ``Offsite radiological impacts of spent nuclear fuel and high-level waste disposal'' issue.... Geologic Repository--Technical Feasibility and Availability C3. Storage of Spent Nuclear Fuel C3.a...

Dry transfer system for spent fuel: Project report, A system designed to achieve the dry transfer of bare spent fuel between two casks. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dawson, D.M.; Guerra, G.; Neider, T.

1995-12-01

This report describes the system developed by EPRI/DOE for the dry transfer of spent fuel assemblies outside the reactor spent fuel pool. The system is designed to allow spent fuel assemblies to be removed from a spent fuel pool in a small cask, transported to the transfer facility, and transferred to a larger cask, either for off-site transportation or on-site storage. With design modifications, this design is capable of transferring single spent fuel assemblies from dry storage casks to transportation casks or visa versa. One incentive for the development of this design is that utilities with limited lifting capacity ormore » other physical or regulatory constraints are limited in their ability to utilize the current, more efficient transportation and storage cask designs. In addition, DOE, in planning to develop and implement the multi-purpose canister (MPC) system for the Civilian Radioactive Waste Management System, included the concept of an on-site dry transfer system to support the implementation of the MPC system at reactors with limitations that preclude the handling of the MPC system transfer casks. This Dry Transfer System can also be used at reactors wi decommissioned spent fuel pools and fuel in dry storage in non-MPC systems to transfer fuel into transportation casks. It can also be used at off-reactor site interim storage facilities for the same purpose.« less

Spent fuel burnup estimation by Cerenkov glow intensity measurement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuribara, Masayuki

1994-10-01

The Cerenkov glow images from irradiated fuel assemblies of boiling-water reactors (BWR) and pressurized-water reactors (PWR) are generally used for inspections. For this purpose, a new UV-I.I. CVD (ultra-violet light image intensifier Cerenkov viewing device), has been developed. This new device can measure the intensity of the Cerenkov glow from a spent fuel assembly, thus making it possible to estimate the burnup of the fuel assembly by comparing the Cerenkov glow intensity to the reference intensity. The experiment was carried out on BWR spent fuel assemblies and the results show that burnups are estimated within 20% accuracy compared to themore » declared burnups for the tested spent fuel assemblies for cooling times ranging from 900--2.000 d.« less

Integral nuclear data validation using experimental spent nuclear fuel compositions

DOE PAGES

Gauld, Ian C.; Williams, Mark L.; Michel-Sendis, Franco; ...

2017-07-19

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors andmore » representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. Furthermore, the database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.« less

Integral nuclear data validation using experimental spent nuclear fuel compositions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauld, Ian C.; Williams, Mark L.; Michel-Sendis, Franco

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors andmore » representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. Furthermore, the database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.« less

Spent fuel container alignment device and method

DOEpatents

Jones, Stewart D.; Chapek, George V.

1996-01-01

An alignment device is used with a spent fuel shipping container including a plurality of fuel pockets for spent fuel arranged in an annular array and having a rotatable cover including an access opening therein. The alignment device includes a lightweight plate which is installed over the access opening of the cover. A laser device is mounted on the plate so as to emit a laser beam through a laser admittance window in the cover into the container in the direction of a pre-established target associated with a particular fuel pocket. An indexing arrangement on the container provides an indication of the angular position of the rotatable cover when the laser beam produced by the laser is brought into alignment with the target of the associated fuel pocket.

10 CFR 72.214 - List of approved spent fuel storage casks.

Code of Federal Regulations, 2010 CFR

2010-01-01

... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.214 List of approved spent...

10 CFR 72.214 - List of approved spent fuel storage casks.

Code of Federal Regulations, 2011 CFR

2011-01-01

... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.214 List of approved spent...

Active Interrogation for Spent Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swinhoe, Martyn Thomas; Dougan, Arden

2015-11-05

The DDA instrument for nuclear safeguards is a fast, non-destructive assay, active neutron interrogation technique using an external 14 MeV DT neutron generator for characterization and verification of spent nuclear fuel assemblies.

Recent developments - US spent fuel disposition

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

One of a US utility's major risk factors in continuing to operate a nuclear plant is managing discharged spent fuel. The US Department of Energy (DOE) signed contracts with utilities guaranteeing government acceptance of spent fuel by 1988. However, on December 17, 1992, DOE Secretary Watkins wrote to Sen. J. Bennett Johnston (D-LA), Chairman of the Senate Energy Committee, indicating a reassessment of DOE's programs, the results of which will be presented to Congress in January 1993. He indicated the Department may not be able to meet the 1988 date, because of difficulty in finding a site for the Monitoredmore » Retrievable Storage facility. Watkins indicated that DOE has investigated an interim solution and decided to expedite a program to certify a multi-purpose standardized cask system for spent fuel receipt, storage, transport, and disposal. To meet the expectations of US utilities, DOE is considering a plan to use federal sites for interim storage of the casks. Secretary Watkins recommended the waste program be taken off-budget and put in a revolving fund established to ensure that money already collected from utilities will be available to meet the schedule for completion of the repository.« less

Spent fuel data base: commercial light water reactors. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

DOE Office of Scientific and Technical Information (OSTI.GOV)

McKinnon, M.A.; Doherty, A.L.

1997-06-01

This report summarizes the results of fuel integrity surveillance determined from gas sampling during and after performance tests and demonstrations conducted from 1983 through 1996 by or in cooperation with the US DOE Office of Commercial Radioactive Waste Management (OCRWM). The cask performance tests were conducted at Idaho National Engineering Laboratory (INEL) between 1984 and 1991 and included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of fuel, and a qualitative determination of the effects of dry storage and fuel consolidation on fission gas release frommore » the spent fuel rods. The performance tests consisted of 6 to 14 runs involving one or two loading, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the end of each performance test, periodic gas sampling was conducted on each cask. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas sampling activities) was initiated by DOE in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are included in this report. Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at INEL offers significant opportunities for confirmation of the benign nature of long-term dry storage. Supporting cask demonstration included licensing and operation of an independent spent fuel storage installation (ISFSI) at the Virginia Power (VP) Surry reactor site. A CASTOR V/21, an MC-10, and a Nuclear Assurance NAC-I28 have been loaded and placed at the VP ISFSI as part of the demonstration program. 13 refs., 14 figs., 9 tabs.« less

Proliferation resistance assessment of various methods of spent nuclear fuel storage and disposal

NASA Astrophysics Data System (ADS)

Kollar, Lenka

Many countries are planning to build or already are building new nuclear power plants to match their growing energy needs. Since all nuclear power plants handle nuclear materials that could potentially be converted and used for nuclear weapons, they each present a nuclear proliferation risk. Spent nuclear fuel presents the largest build-up of nuclear material at a power plant. This is a proliferation risk because spent fuel contains plutonium that can be chemically separated and used for a nuclear weapon. The International Atomic Energy Agency (IAEA) safeguards spent fuel in all non-nuclear weapons states that are party to the Non-Proliferation Treaty. Various safeguards methods are in use at nuclear power plants and research is underway to develop safeguards methods for spent fuel in centralized storage or underground storage and disposal. Each method of spent fuel storage presents different proliferation risks due to the nature of the storage method and the safeguards techniques that are utilized. Previous proliferation resistance and proliferation risk assessments have mainly compared nuclear material through the whole fuel cycle and not specifically focused on spent fuel storage. This project evaluates the proliferation resistance of the three main types of spent fuel storage: spent fuel pool, dry cask storage, and geological repository. The proliferation resistance assessment methodology that is used in this project is adopted from previous work and altered to be applicable to spent fuel storage. The assessment methodology utilizes various intrinsic and extrinsic proliferation-resistant attributes for each spent fuel storage type. These attributes are used to calculate a total proliferation resistant (PR) value. The maximum PR value is 1.00 and a greater number means that the facility is more proliferation resistant. Current data for spent fuel storage in the United States and around the world was collected. The PR values obtained from this data are 0.49 for

78 FR 66858 - Waste Confidence-Continued Storage of Spent Nuclear Fuel

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-07

...-2012-0246] RIN 3150-AJ20 Waste Confidence--Continued Storage of Spent Nuclear Fuel AGENCY: Nuclear... its generic determination on the environmental impacts of the continued storage of spent nuclear fuel... revising the generic determination of the environmental impacts of the continued storage of spent nuclear...

Spent fuel treatment at ANL-West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goff, K.M.; Benedict, R.W.; Levinskas, D.

1994-12-31

At Argonne National Laboratory-West (ANL-West) there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Cycle Facility at ANL-West to produce stable waste forms for storage and disposal. The treatment operations will employ a pyrochemical process that also has applications for treating most of the fuel types within the Department of Energy complex. The treatment equipment is in its last stage of readiness, and operations will begin in the Fall of 1994.

Optimization of spent fuel pool weir gate driving mechanism

NASA Astrophysics Data System (ADS)

Liu, Chao; Du, Lin; Tao, Xinlei; Wang, Shijie; Shang, Ertao; Yu, Jianjiang

2018-04-01

Spent fuel pool is crucial facility for fuel storage and nuclear safety, and the spent fuel pool weir gate is the key related equipment. In order to achieve a goal of more efficient driving force transfer, loading during the opening/closing process is analyzed and an optimized calculation method for dimensions of driving mechanism is proposed. The result of optimizing example shows that the method can be applied to weir gates' design with similar driving mechanism.

MTR plates modeling with MAIA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marelle, V.; Dubois, S.; Ripert, M.

2008-07-15

MAIA is a thermo-mechanical code dedicated to the modeling of MTR fuel plates. The main physical phenomena modeled in the code are the cladding oxidation, the interaction between fuel and Al-matrix, the swelling due to fission products and the Al/fuel particles interaction. The creeping of the plate can be modeled in the mechanical calculation. MAIA has been validated on U-Mo dispersion fuel experiments such as IRIS 1 and 2 and FUTURE. The results are in rather good agreement with post-irradiation examinations. MAIA can also be used to calculate in-pile behavior of U{sub 3}Si{sub 2} plates as in the SHARE experimentmore » irradiated in the SCK/Mol BR2 reactor. The main outputs given by MAIA throughout the irradiation are temperatures, cladding oxidation thickness, interaction thickness, volume fraction of meat constituents, swelling, displacements, strains and stresses. MAIA is originally a two-dimensional code but a three-dimensional version is currently under development. (author)« less

Array Detector Modules for Spent Fuel Verification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bolotnikov, Aleksey

Brookhaven National Laboratory (BNL) proposes to evaluate the arrays of position-sensitive virtual Frisch-grid (VFG) detectors for passive gamma-ray emission tomography (ET) to verify the spent fuel in storage casks before storing them in geo-repositories. Our primary objective is to conduct a preliminary analysis of the arrays capabilities and to perform field measurements to validate the effectiveness of the proposed array modules. The outcome of this proposal will consist of baseline designs for the future ET system which can ultimately be used together with neutrons detectors. This will demonstrate the usage of this technology in spent fuel storage casks.

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

NASA Astrophysics Data System (ADS)

Margeanu, C. A.; Margeanu, S.; Barbos, D.; Iorgulis, C.

2010-12-01

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

Spent Fuel Ratio Estimates from Numerical Models in ALE3D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Margraf, J. D.; Dunn, T. A.

Potential threat of intentional sabotage of spent nuclear fuel storage facilities is of significant importance to national security. Paramount is the study of focused energy attacks on these materials and the potential release of aerosolized hazardous particulates into the environment. Depleted uranium oxide (DUO 2) is often chosen as a surrogate material for testing due to the unreasonable cost and safety demands for conducting full-scale tests with real spent nuclear fuel. To account for differences in mechanical response resulting in changes to particle distribution it is necessary to scale the DUO 2 results to get a proper measure for spentmore » fuel. This is accomplished with the spent fuel ratio (SFR), the ratio of respirable aerosol mass released due to identical damage conditions between a spent fuel and a surrogate material like depleted uranium oxide (DUO 2). A very limited number of full-scale experiments have been carried out to capture this data, and the oft-questioned validity of the results typically leads to overly-conservative risk estimates. In the present work, the ALE3D hydrocode is used to simulate DUO 2 and spent nuclear fuel pellets impacted by metal jets. The results demonstrate an alternative approach to estimate the respirable release fraction of fragmented nuclear fuel.« less

Detecting pin diversion from pressurized water reactors spent fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ham, Young S.; Sitaraman, Shivakumar

Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and takingmore » the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.« less

Systems for the Intermodal Routing of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peterson, Steven K; Liu, Cheng

The safe and secure movement of spent nuclear fuel from shutdown and active reactor facilities to intermediate or long term storage sites may, in some instances, require the use of several modes of transportation to accomplish the move. To that end, a fully operable multi-modal routing system is being developed within Oak Ridge National Laboratory s (ORNL) WebTRAGIS (Transportation Routing Analysis Geographic Information System). This study aims to provide an overview of multi-modal routing, the existing state of the TRAGIS networks, the source data needs, and the requirements for developing structural relationships between various modes to create a suitable systemmore » for modeling the transport of spent nuclear fuel via a multimodal network. Modern transportation systems are comprised of interconnected, yet separate, modal networks. Efficient transportation networks rely upon the smooth transfer of cargoes at junction points that serve as connectors between modes. A key logistical impediment to the shipment of spent nuclear fuel is the absence of identified or designated transfer locations between transport modes. Understanding the potential network impacts on intermodal transportation of spent nuclear fuel is vital for planning transportation routes from origin to destination. By identifying key locations where modes intersect, routing decisions can be made to prioritize cost savings, optimize transport times and minimize potential risks to the population and environment. In order to facilitate such a process, ORNL began the development of a base intermodal network and associated routing code. The network was developed using previous intermodal networks and information from publicly available data sources to construct a database of potential intermodal transfer locations with likely capability to handle spent nuclear fuel casks. The coding development focused on modifying the existing WebTRAGIS routing code to accommodate intermodal transfers and the

Advancing the Fork detector for quantitative spent nuclear fuel verification

DOE PAGES

Vaccaro, S.; Gauld, I. C.; Hu, J.; ...

2018-01-31

The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations.more » A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This study describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false

Advancing the Fork detector for quantitative spent nuclear fuel verification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaccaro, S.; Gauld, I. C.; Hu, J.

The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations.more » A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This study describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false

Advancing the Fork detector for quantitative spent nuclear fuel verification

NASA Astrophysics Data System (ADS)

Vaccaro, S.; Gauld, I. C.; Hu, J.; De Baere, P.; Peterson, J.; Schwalbach, P.; Smejkal, A.; Tomanin, A.; Sjöland, A.; Tobin, S.; Wiarda, D.

2018-04-01

The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations. A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This paper describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false positive alarms

MTR BUILDING, TRA603. DETAILED VIEW OF NORTHWEST CORNERS OF MTR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR BUILDING, TRA-603. DETAILED VIEW OF NORTHWEST CORNERS OF MTR HIGH-BAY AND SECOND/THIRD STORY SECTIONS. NOTE SHAPE OF PANEL ABOVE WINDOW OVER "TRA-603" BUILDING NUMBERS. THIS IS A "STANDARD PANEL." INL NEGATIVE NUMBER HD46-42-2. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

78 FR 32077 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-29

... Fuel Storage Casks: MAGNASTOR[supreg] System AGENCY: Nuclear Regulatory Commission. ACTION: Direct... final rule that would have revised its spent fuel storage regulations to include Amendment No. 3 to... All-purpose Storage (MAGNASTOR[supreg]) System listing within the ``List of Approved Spent Fuel...

Development of Techniques for Spent Fuel Assay – Differential Dieaway Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swinhoe, Martyn Thomas; Goodsell, Alison; Ianakiev, Kiril Dimitrov

This report summarizes the work done under a DNDO R&D funded project on the development of the differential dieaway method to measure plutonium in spent fuel. There are large amounts of plutonium that are contained in spent fuel assemblies, and currently there is no way to make quantitative non-destructive assay. This has led NA24 under the Next Generation Safeguards Initiative (NGSI) to establish a multi-year program to investigate, develop and implement measurement techniques for spent fuel. The techniques which are being experimentally tested by the existing NGSI project do not include any pulsed neutron active techniques. The present work coversmore » the active neutron differential dieaway technique and has advanced the state of knowledge of this technique as well as produced a design for a practical active neutron interrogation instrument for spent fuel. Monte Carlo results from the NGSI effort show that much higher accuracy (1-2%) for the Pu content in spent fuel assemblies can be obtained with active neutron interrogation techniques than passive techniques, and this would allow their use for nuclear material accountancy independently of any information from the operator. The main purpose of this work was to develop an active neutron interrogation technique for spent nuclear fuel.« less

Pyroprocess for processing spent nuclear fuel

DOEpatents

Miller, William E.; Tomczuk, Zygmunt

2002-01-01

This is a pyroprocess for processing spent nuclear fuel. The spent nuclear fuel is chopped into pieces and placed in a basket which is lowered in to a liquid salt solution. The salt is rich in ZrF.sub.4 and containing alkali or alkaline earth fluorides, and in particular, the salt chosen was LiF-50 mol % ZrF.sub.4 with a eutectic melting point of 500.degree. C. Prior to lowering the basket, the salt is heated to a temperature of between 550.degree. C. and 700.degree. C. in order to obtain a molten solution. After dissolution the oxides of U, Th, rare earth and other like oxides, the salt bath solution is subject to hydro-fluorination to remove the oxygen and then to a fluorination step to remove U as gaseous UF.sub.6. In addition, after dissolution, the basket contains PuO.sub.2 and undissolved parts of the fuel rods, and the basket and its contents are processed to remove the Pu.

Fiscal Year (FY) 2017 Activities for the Spent Fuel Nondestructive Assay Project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trellue, Holly Renee; Trahan, Alexis Chanel; McMath, Garrett Earl

The main focus of research in the NA-241 spent fuel nondestructive assay (NDA) project in FY17 has been completing the fabrication and testing of two prototype instruments for upcoming spent fuel measurements at the Clab interim storage facility in Sweden. One is a passive instrument: Differential Die-away Self Interrogation-Passive Neutron Albedo Reactivity (DDSI), and one is an active instrument: Differential Die-Away-Californium Interrogation with Prompt Neutron (DDA). DDSI was fabricated and tested with fresh fuel at Los Alamos National Laboratory in FY15 and FY16, then shipped to Sweden at the beginning of FY17. Research was performed in FY17 to simplify resultsmore » from the data acquisition system, which is complex because signals from 56 different 3He detectors must be processed using list mode data. The DDA instrument was fabricated at the end of FY16. New high count rate electronics better suited for a spent fuel environment (i.e., KM-200 preamplifiers) were built specifically for this instrument in FY17, and new Tygon tubing to house electrical cables was purchased and installed. Fresh fuel tests using the DDA instrument with numerous configurations of fuel rods containing depleted uranium (DU), low enriched uranium (LEU), and LEU with burnable poisons (Gd) were successfully performed and compared to simulations.1 Additionally, members of the spent fuel NDA project team travelled to Sweden for a “spent fuel characterization and decay heat” workshop involving simulations of spent fuel and analysis of uncertainties in decay heat calculations.« less

Pyroprocessing of Light Water Reactor Spent Fuels Based on an Electrochemical Reduction Technology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ohta, Hirokazu; Inoue, Tadashi; Sakamura, Yoshiharu

A concept of pyroprocessing light water reactor (LWR) spent fuels based on an electrochemical reduction technology is proposed, and the material balance of the processing of mixed oxide (MOX) or high-burnup uranium oxide (UO{sub 2}) spent fuel is evaluated. Furthermore, a burnup analysis for metal fuel fast breeder reactors (FBRs) is conducted on low-decontamination materials recovered by pyroprocessing. In the case of processing MOX spent fuel (40 GWd/t), UO{sub 2} is separately collected for {approx}60 wt% of the spent fuel in advance of the electrochemical reduction step, and the product recovered through the rare earth (RE) removal step, which hasmore » the composition uranium:plutonium:minor actinides:fission products (FPs) = 76.4:18.4:1.7:3.5, can be applied as an ingredient of FBR metal fuel without a further decontamination process. On the other hand, the electroreduced alloy of high-burnup UO{sub 2} spent fuel (48 GWd/t) requires further decontamination of residual FPs by an additional process such as electrorefining even if RE FPs are removed from the alloy because the recovered plutonium (Pu) is accompanied by almost the same amount of FPs in addition to RE. However, the amount of treated materials in the electrorefining step is reduced to {approx}10 wt% of the total spent fuel owing to the prior UO{sub 2} recovery step. These results reveal that the application of electrochemical reduction technology to LWR spent oxide fuel is a promising concept for providing FBR metal fuel by a rationalized process.« less

Analysis of Transportation Options for Commercial Spent Fuel in the U.S.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalinina, Elena; Busch, Ingrid Karin

.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage and associated transportation of spent nuclear fuel (SNF) highand associated transportation of spent nuclear fuel (SNF) and high and

MTR MAIN FLOOR. NEUTRON TUNNEL (SPANNED BY STILELIKE STEPS) PROJECTS ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR MAIN FLOOR. NEUTRON TUNNEL (SPANNED BY STILE-LIKE STEPS) PROJECTS FROM THE SOUTHEAST CORNER OF THE MTR TOWARD SOUTHEAST CORNER OF BUILDING, WHERE SHIELDING BLOCKS BEGIN TO SURROUND THE TUNNEL AS IT NEARS DETECTING INSTRUMENTS NEAR THE BUILDING WALL. GEAR RELATED TO CRYSTAL NEUTRON SPECTROMETER IS IN FOREGROUND SURROUNDED BY SHIELDING. DATA CONSOLES ARE AT MID-LEVEL OF EAST FACE. OTHER WORK PROCEEDS ON TOP OF AND ELSEWHERE AROUND REACTOR. NOTE TOOLS HANGING AGAINST SOUTHEAST CORNER, USED TO CHANGE FUEL ELEMENTS AND OTHER REACTOR ITEMS DURING REFUELING CYCLES. INL NEGATIVE NO. 10439. Unknown Photographer, 4/20/1954 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

78 FR 40199 - Draft Spent Fuel Storage and Transportation Interim Staff Guidance

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-03

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0140] Draft Spent Fuel Storage and Transportation Interim... Spent Fuel Storage and Transportation Interim Staff Guidance No. 24 (SFST-ISG-24), Revision 0, ``The Use of a Demonstration Program as Confirmation of Integrity for Continued Storage of High Burnup Fuel...

Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement

DOE Office of Scientific and Technical Information (OSTI.GOV)

N /A

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

Air Shipment of Spent Nuclear Fuel from Romania to Russia

DOE Office of Scientific and Technical Information (OSTI.GOV)

Igor Bolshinsky; Ken Allen; Lucian Biro

Romania successfully completed the world’s first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities formore » shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.« less

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

DOEpatents

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

2013-11-05

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

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

DOEpatents

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

2014-06-10

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

Hanford Spent Nuclear Fuel Project recommended path forward

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fulton, J.C.

The Spent Nuclear Fuel Project (the Project), in conjunction with the U.S. Department of Energy-commissioned Independent Technical Assessment (ITA) team, has developed engineered alternatives for expedited removal of spent nuclear fuel, including sludge, from the K Basins at Hanford. These alternatives, along with a foreign processing alternative offered by British Nuclear Fuels Limited (BNFL), were extensively reviewed and evaluated. Based on these evaluations, a Westinghouse Hanford Company (WHC) Recommended Path Forward for K Basins spent nuclear fuel has been developed and is presented in Volume I of this document. The recommendation constitutes an aggressive series of projects to construct andmore » operate systems and facilities to safely retrieve, package, transport, process, and store K Basins fuel and sludge. The overall processing and storage scheme is based on the ITA team`s proposed passivation and vault storage process. A dual purpose staging and vault storage facility provides an innovative feature which allows accelerated removal of fuel and sludge from the basins and minimizes programmatic risks beyond any of the originally proposed alternatives. The projects fit within a regulatory and National Environmental Policy Act (NEPA) overlay which mandates a two-phased approach to construction and operation of the needed facilities. The two-phase strategy packages and moves K Basins fuel and sludge to a newly constructed Staging and Storage Facility by the year 2000 where it is staged for processing. When an adjoining facility is constructed, the fuel is cycled through a stabilization process and returned to the Staging and Storage Facility for dry interim (40-year) storage. The estimated total expenditure for this Recommended Path Forward, including necessary new construction, operations, and deactivation of Project facilities through 2012, is approximately $1,150 million (unescalated).« less

10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Procedures for spent fuel storage cask submittals. 72.230 Section 72.230 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

10 CFR 72.240 - Conditions for spent fuel storage cask reapproval.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Conditions for spent fuel storage cask reapproval. 72.240 Section 72.240 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

Spent fuel behavior under abnormal thermal transients during dry storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stahl, D.; Landow, M.P.; Burian, R.J.

1986-01-01

This study was performed to determine the effects of abnormally high temperatures on spent fuel behavior. Prior to testing, calculations using the CIRFI3 code were used to determine the steady-state fuel and cask component temperatures. The TRUMP code was used to determine transient heating rates under postulated abnormal events during which convection cooling of the cask surfaces was obstructed by a debris bed covering the cask. The peak rate of temperature rise during the first 6 h was calculated to be about 15/sup 0/C/h, followed by a rate of about 1/sup 0/C/h. A Turkey Point spent fuel rod segment wasmore » heated to approx. 800/sup 0/C. The segment deformed uniformly with an average strain of 17% at failure and a local strain of 60%. Pretest characterization of the spent fuel consisted of visual examination, profilometry, eddy-current examination, gamma scanning, fission gas collection, void volume measurement, fission gas analysis, hydrogen analysis of the cladding, burnup analysis, cladding metallography, and fuel ceramography. Post-test characterization showed that the failure was a pinhole cladding breach. The results of the tests showed that spent fuel temperatures in excess of 700/sup 0/C are required to produce a cladding breach in fuel rods pressurized to 500 psing (3.45 MPa) under postulated abnormal thermal transient cask conditions. The pinhole cladding breach that developed would be too small to compromise the confinement of spent fuel particles during an abnormal event or after normal cooling conditions are restored. This behavior is similar to that found in other slow ramp tests with irradiated and nonirradiated rod sections and nonirradiated whole rods under conditions that bracketed postulated abnormal heating rates. This similarity is attributed to annealing of the irradiation-strengthened Zircaloy cladding during heating. In both cases, the failure was a benign, ductile pinhole rupture.« less

Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

NASA Astrophysics Data System (ADS)

Acır, Adem; Altunok, Taner

2010-10-01

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li20Sn80 coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.

Gamma-ray mirror technology for NDA of spent fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Descalle, M. A.; Ruz-Armendariz, J.; Decker, T.

Direct measurements of gamma rays emitted by fissile material have been proposed as an alternative to measurements of the gamma rays from fission products. From a safeguards applications perspective, direct detection of uranium (U) and plutonium (Pu) K-shell fluorescence emission lines and specific lines from some of their isotopes could lead to improved shipper-receiver difference or input accountability at the start of Pu reprocessing. However, these measurements are difficult to implement when the spent fuel is in the line-of-sight of the detector, as the detector is exposed to high rates dominated by fission product emissions. To overcome the combination ofmore » high rates and high background, grazing incidence multilayer mirrors have been proposed as a solution to selectively reflect U and Pu hard X-ray and soft gamma rays in the 90 to 420 keV energy into a high-purity germanium (HPGe) detector shielded from the direct line-of-sight of spent fuel. Several groups demonstrated that K-shell fluorescence lines of U and Pu in spent fuel could be detected with Ge detectors. In the field of hard X-ray optics the performance of reflective multilayer coated reflective optics was demonstrated up to 645 keV at the European Synchrotron Radiation Facility. Initial measurements conducted at Oak Ridge National Laboratory with sealed sources and scoping experiments conducted at the ORNL Irradiated Fuels Examination Laboratory (IFEL) with spent nuclear fuel further demonstrated the pass-band properties of multilayer mirrors for reflecting specific emission lines into 1D and 2D HPGe detectors, respectively.« less

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

NASA Astrophysics Data System (ADS)

Wiss, Thierry; Hiernaut, Jean-Pol; Roudil, Danièle; Colle, Jean-Yves; Maugeri, Emilio; Talip, Zeynep; Janssen, Arne; Rondinella, Vincenzo; Konings, Rudy J. M.; Matzke, Hans-Joachim; Weber, William J.

2014-08-01

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

Estimating the time for dissolution of spent fuel exposed to unlimited water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leider, H.R.; Nguyen, S.N.; Stout, R.B.

1991-12-01

The release of radionuclides from spent fuel cannot be precisely predicted at this point because a satisfactory dissolution model based on specific chemical processes is not yet available. However, preliminary results on the dissolution rate of UO{sub 2} and spent fuel as a function of temperature and water composition have recently been reported. This information, together with data on fragment size distribution of spent fuel, are used to estimate the dissolution response of spent fuel in excess flowing water within the framework of a simple model. In this model, the reaction/dissolution front advances linearly with time and geometry is preserved.more » This also estimates the dissolution rate of the bulk of the fission products and higher actinides, which are uniformly distributed in the UO{sub 2} matrix and are presumed to dissolve congruently. We have used a fuel fragment distribution actually observed to calculate the time for total dissolution of spent fuel. A worst-case estimate was also made using the initial (maximum) rate of dissolution to predict the total dissolution time. The time for total dissolution of centimeter size particles is estimated to be 5.5 {times} 10{sup 4} years at 25{degrees}C.« less

Characterization of neutron sources from spent fuel casks. [Skyshine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parks, C.V.; Pace, J.V. III

1987-01-01

In the interim period prior to the acceptance of spent fuel for disposal by the USDOE, utilities are beginning to choose dry cask storage as an alternative to pool re-racking, transshipments, or new pool construction. In addition, the current MRS proposal calls for interim dry storage of consolidated spent fuel in concrete casks. As part of the licensing requirements for these cask storage facilities, calculations are typically necessary to determine the yearly radiation dose received at the site boundary. Unlike wet facilities, neutron skyshine can be an important contribution to the total boundary dose from a dry storage facility. Calculationmore » of the neutron skyshine is in turn heavily dependent on the source characteristics and source model selected for the analysis. This paper presents the basic source characteristics of the spent fuel stored in dry casks and discusses factors that must be considered in evaluating and modeling the radiation sources for the subsequent skyshine calculation. 4 refs., 1 tab.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Christopher Landers; Igor Bolshinsky; Ken Allen

2010-07-01

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

In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauld, Ian C.; Hu, Jianwei; De Baere, P.

Expanding spent fuel dry storage activities worldwide are increasing demands on safeguards authorities that perform inspections. The European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) require measurements to verify declarations when spent fuel is transferred to difficult-to-access locations, such as dry storage casks and the repositories planned in Finland and Sweden. EURATOM makes routine use of the Fork detector to obtain gross gamma and total neutron measurements during spent fuel inspections. Data analysis is performed by modules in the integrated Review and Analysis Program (iRAP) software, developed jointly by EURATOM and the IAEA. Under the frameworkmore » of the US Department of Energy–EURATOM cooperation agreement, a module for automated Fork detector data analysis has been developed by Oak Ridge National Laboratory (ORNL) using the ORIGEN code from the SCALE code system and implemented in iRAP. EURATOM and ORNL recently performed measurements on 30 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab), operated by the Swedish Nuclear Fuel and Waste Management Company (SKB). The measured assemblies represent a broad range of fuel characteristics. Neutron count rates for 15 measured pressurized water reactor assemblies are predicted with an average relative standard deviation of 4.6%, and gamma signals are predicted on average within 2.6% of the measurement. The 15 measured boiling water reactor assemblies exhibit slightly larger deviations of 5.2% for the gamma signals and 5.7% for the neutron count rates, compared to measurements. These findings suggest that with improved analysis of the measurement data, existing instruments can provide increased verification of operator declarations of the spent fuel and thereby also provide greater ability to confirm integrity of an assembly. These results support the application of the Fork detector as a fully quantitative

PWR and BWR spent fuel assembly gamma spectra measurements

NASA Astrophysics Data System (ADS)

Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

2016-10-01

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

PWR and BWR spent fuel assembly gamma spectra measurements

DOE PAGES

Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; ...

2016-07-17

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

PWR and BWR spent fuel assembly gamma spectra measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

Safeguards-by-Design: Guidance for Independent Spent Fuel Dry Storage Installations (ISFSI)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trond Bjornard; Philip C. Durst

2012-05-01

This document summarizes the requirements and best practices for implementing international nuclear safeguards at independent spent fuel storage installations (ISFSIs), also known as Away-from- Reactor (AFR) storage facilities. These installations may provide wet or dry storage of spent fuel, although the safeguards guidance herein focuses on dry storage facilities. In principle, the safeguards guidance applies to both wet and dry storage. The reason for focusing on dry independent spent fuel storage installations is that this is one of the fastest growing nuclear installations worldwide. Independent spent fuel storage installations are typically outside of the safeguards nuclear material balance area (MBA)more » of the reactor. They may be located on the reactor site, but are generally considered by the International Atomic Energy Agency (IAEA) and the State Regulator/SSAC to be a separate facility. The need for this guidance is becoming increasingly urgent as more and more nuclear power plants move their spent fuel from resident spent fuel ponds to independent spent fuel storage installations. The safeguards requirements and best practices described herein are also relevant to the design and construction of regional independent spent fuel storage installations that nuclear power plant operators are starting to consider in the absence of a national long-term geological spent fuel repository. The following document has been prepared in support of two of the three foundational pillars for implementing Safeguards-by-Design (SBD). These are: i) defining the relevant safeguards requirements, and ii) defining the best practices for meeting the requirements. This document was prepared with the design of the latest independent dry spent fuel storage installations in mind and was prepared specifically as an aid for designers of commercial nuclear facilities to help them understand the relevant international requirements that follow from a country’s safeguards agreement

10 CFR 72.240 - Conditions for spent fuel storage cask reapproval.

Code of Federal Regulations, 2011 CFR

2011-01-01

... has been determined by the NRC. The application must be accompanied by a safety analysis report (SAR). The new SAR may reference the SAR originally submitted for the approved spent fuel storage cask design. (c) The design of a spent fuel storage cask will be reapproved if the conditions in § 72.238 are met...

REACTIVITY MEASUREMENT FACILITY. CAMERA LOOKS DOWN INTO MTR CANAL. REACTOR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

REACTIVITY MEASUREMENT FACILITY. CAMERA LOOKS DOWN INTO MTR CANAL. REACTOR IS FUELED AS AN ETR MOCK-UP. LIGHTS DANGLE BELOW WATER LEVEL. CONTROL RODS AND OTHER APPARATUS DESCEND FROM ABOVE WATER LEVEL. INL NEGATIVE NO. 56-900. Jack L. Anderson, Photographer, 3/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

The Role of the Transcription Factors MtrR and MtrA in the Fitness of the Pathogen Neisseria gonorrhoeae

DTIC Science & Technology

2007-10-19

MacA -MacB efflux pump is expressed at very low levels in N. gonorrhoeae; however, experimental alteration of the promoter sequence showed that the...specificity (132). The gonococcus expresses four efflux pump systems, namely MtrC-Mtr-D-MtrE (64), FarA-FarB-MtrE (97), NorM (158), and MacA -MacB (161...2005. Characterization of the MacA -MacB efflux system in Neisseria gonorrhoeae. The Journal of antimicrobial chemotherapy 56:856-860. 162. Rouquette, C

Neutron radiation characteristics of the IVth generation reactor spent fuel

NASA Astrophysics Data System (ADS)

Bedenko, Sergey; Shamanin, Igor; Grachev, Victor; Knyshev, Vladimir; Ukrainets, Olesya; Zorkin, Andrey

2018-03-01

Exploitation of nuclear power plants as well as construction of new generation reactors lead to great accumulation of spent fuel in interim storage facilities at nuclear power plants, and in spent fuel «wet» and «dry» long-term storages. Consequently, handling the fuel needs more attention. The paper is focused on the creation of an efficient computational model used for developing the procedures and regulations of spent nuclear fuel handling in nuclear fuel cycle of the new generation reactor. A Thorium High-temperature Gas-Cooled Reactor Unit (HGTRU, Russia) was used as an object for numerical research. Fuel isotopic composition of HGTRU was calculated using the verified code of the MCU-5 program. The analysis of alpha emitters and neutron radiation sources was made. The neutron yield resulting from (α,n)-reactions and at spontaneous fission was calculated. In this work it has been shown that contribution of (α,n)-neutrons is insignificant in case of such (Th,Pu)-fuel composition and HGTRU operation mode, and integral neutron yield can be approximated by the Watt spectral function. Spectral and standardized neutron distributions were achieved by approximation of the list of high-precision nuclear data. The distribution functions were prepared in group and continuous form for further use in calculations according to MNCP, MCU, and SCALE.

78 FR 16619 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-18

...-0308] RIN 3150-AJ22 List of Approved Spent Fuel Storage Casks: MAGNASTOR[supreg] System AGENCY: Nuclear... proposing to amend its spent fuel storage regulations by revising the NAC International, Inc., Modular Advanced Generation Nuclear All-purpose Storage (MAGNASTOR[supreg]) Cask System listing within the ``List...

Spent Fuel Working Group Report. Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

O`Toole, T.

1993-11-01

The Department of Energy is storing large amounts of spent nuclear fuel and other reactor irradiated nuclear materials (herein referred to as RINM). In the past, the Department reprocessed RINM to recover plutonium, tritium, and other isotopes. However, the Department has ceased or is phasing out reprocessing operations. As a consequence, Department facilities designed, constructed, and operated to store RINM for relatively short periods of time now store RINM, pending decisions on the disposition of these materials. The extended use of the facilities, combined with their known degradation and that of their stored materials, has led to uncertainties about safety.more » To ensure that extended storage is safe (i.e., that protection exists for workers, the public, and the environment), the conditions of these storage facilities had to be assessed. The compelling need for such an assessment led to the Secretary`s initiative on spent fuel, which is the subject of this report. This report comprises three volumes: Volume I; Summary Results of the Spent Fuel Working Group Evaluation; Volume II, Working Group Assessment Team Reports and Protocol; Volume III; Operating Contractor Site Team Reports. This volume presents the overall results of the Working Group`s Evaluation. The group assessed 66 facilities spread across 11 sites. It identified: (1) facilities that should be considered for priority attention. (2) programmatic issues to be considered in decision making about interim storage plans and (3) specific vulnerabilities for some of these facilities.« less

MATERIALS TESTING REACTOR (MTR) BUILDING, TRA603. CONTEXTUAL VIEW OF MTR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MATERIALS TESTING REACTOR (MTR) BUILDING, TRA-603. CONTEXTUAL VIEW OF MTR BUILDING SHOWING NORTH SIDES OF THE HIGH-BAY REACTOR BUILDING, ITS SECOND/THIRD FLOOR BALCONY LEVEL, AND THE ATTACHED ONE-STORY OFFICE/LABORATORY BUILDING, TRA-604. CAMERA FACING SOUTHEAST. VERTICAL CONCRETE-SHROUDED BEAMS SUPPORT PRECAST CONCRETE PANELS. CONCRETE PROJECTION FORMED AS A BUNKER AT LEFT OF VIEW IS TRA-657, PLUG STORAGE BUILDING. INL NEGATIVE NO. HD46-42-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Quantity and management of spent fuel from prototype and research reactors in Germany

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dorr, Sabine; Bollingerfehr, Wilhelm; Filbert, Wolfgang

Within the scope of an R and D project (project identification number FKZ 02 S 8679) sponsored by BMBF (Federal Ministry of Education and Research), the current state of storage and management of fuel elements from prototype and research reactors was established, and an approach for their future storage/management was developed. The spent fuels from prototype and research reactors in Germany that require disposal were specified and were described in regard to their repository-relevant characteristics. As there are currently no casks licensed for disposal in Germany, descriptions of casks that were considered to be suitable were provided. Based on themore » information provided on the spent fuel from prototype and research reactors and the potential casks, a technical disposal concept was developed. In this context, concepts to integrate the spent fuel from prototype and research reactors into existing disposal concepts for spent fuel from German nuclear power plants and for waste from reprocessing were developed for salt and clay formations. (authors)« less

Dissolution of spent nuclear fuel in carbonate-peroxide solution

NASA Astrophysics Data System (ADS)

Soderquist, Chuck; Hanson, Brady

2010-01-01

This study shows that spent UO2 fuel can be completely dissolved in a room temperature carbonate-peroxide solution apparently without attacking the metallic Mo-Tc-Ru-Rh-Pd fission product phase. In parallel tests, identical samples of spent nuclear fuel were dissolved in nitric acid and in an ammonium carbonate, hydrogen peroxide solution. The resulting solutions were analyzed for strontium-90, technetium-99, cesium-137, europium-154, plutonium, and americium-241. The results were identical for all analytes except technetium, where the carbonate-peroxide dissolution had only about 25% of the technetium that the nitric acid dissolution had.

An integrated approach for determining plutonium mass in spent fuel assemblies with nondestructive assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swinhoe, Martyn T; Tobin, Stephen J; Fensin, Mike L

2009-01-01

There are a variety of reasons for quantifying plutonium (Pu) in spent fuel. Below, five motivations are listed: (1) To verify the Pu content of spent fuel without depending on unverified information from the facility, as requested by the IAEA ('independent verification'). New spent fuel measurement techniques have the potential to allow the IAEA to recover continuity of knowledge and to better detect diversion. (2) To assure regulators that all of the nuclear material of interest leaving a nuclear facility actually arrives at another nuclear facility ('shipper/receiver'). Given the large stockpile of nuclear fuel at reactor sites around the world,more » it is clear that in the coming decades, spent fuel will need to be moved to either reprocessing facilities or storage sites. Safeguarding this transportation is of significant interest. (3) To quantify the Pu in spent fuel that is not considered 'self-protecting.' Fuel is considered self-protecting by some regulatory bodies when the dose that the fuel emits is above a given level. If the fuel is not self-protecting, then the Pu content of the fuel needs to be determined and the Pu mass recorded in the facility's accounting system. This subject area is of particular interest to facilities that have research-reactor spent fuel or old light-water reactor (LWR) fuel. It is also of interest to regulators considering changing the level at which fuel is considered self-protecting. (4) To determine the input accountability value at an electrochemical processing facility. It is not expected that an electrochemical reprocessing facility will have an input accountability tank, as is typical in an aqueous reprocessing facility. As such, one possible means of determining the input accountability value is to measure the Pu content in the spent fuel that arrives at the facility. (5) To fully understand the composition of the fuel in order to efficiently and safely pack spent fuel into a long-term repository. The NDA of spent

Antineutrino Monitoring of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Brdar, Vedran; Huber, Patrick; Kopp, Joachim

2017-11-01

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

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

DOE PAGES

Hu, Jianwei; Giaquinto, J. M.; Gauld, I. C.; ...

2017-04-28

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor havemore » been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.« less

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Giaquinto, J. M.; Gauld, I. C.

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor havemore » been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly Cummins; Igor Bolshinsky; Ken Allen

2009-07-01

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

Spent fuel cask handling at an operating nuclear power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pal, A.C.

1988-01-01

The importance of spent fuel handling at operating nuclear power plants cannot be overstated. Because of its highly radioactive nature, however, spent fuel must be handled in thick, lead-lined containers or casks. Thus, all casks for spent fuel handling are heavy loads by the US Nuclear Regulatory Commission's definition, and any load-drop must be evaluated for its potential to damage safety-related equipment. Nuclear Regulatory Guide NUREG-0612 prescribes the regulatory requirements of alternative heavy-load-handling methodologies such as (a) by providing cranes that meet the requirements of NUREG-0554, which shall be called the soft path, or (b) by providing protective devices atmore » all postulated load-drop areas to prevent any damage to safety-related equipment, which shall be called the hard path. The work reported in this paper relates to cask handling at New York Power Authority's James A. FitzPatrick (JAF) plant.« less

Molten tin reprocessing of spent nuclear fuel elements

DOEpatents

Heckman, Richard A.

1983-01-01

A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

Application of Compton-suppressed self-induced XRF to spent nuclear fuel measurement

NASA Astrophysics Data System (ADS)

Park, Se-Hwan; Jo, Kwang Ho; Lee, Seung Kyu; Seo, Hee; Lee, Chaehun; Won, Byung-Hee; Ahn, Seong-Kyu; Ku, Jeong-Hoe

2017-11-01

Self-induced X-ray fluorescence (XRF) is a technique by which plutonium (Pu) content in spent nuclear fuel can be directly quantified. In the present work, this method successfully measured the plutonium/uranium (Pu/U) peak ratio of a pressurized water reactor (PWR)'s spent nuclear fuel at the Korea atomic energy research institute (KAERI)'s post irradiation examination facility (PIEF). In order to reduce the Compton background in the low-energy X-ray region, the Compton suppression system additionally was implemented. By use of this system, the spectrum's background level was reduced by a factor of approximately 2. This work shows that Compton-suppressed selfinduced XRF can be effectively applied to Pu accounting in spent nuclear fuel.

10 CFR 171.15 - Annual fees: Reactor licenses and independent spent fuel storage licenses.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Annual fees: Reactor licenses and independent spent fuel... REACTOR LICENSES AND FUEL CYCLE LICENSES AND MATERIALS LICENSES, INCLUDING HOLDERS OF CERTIFICATES OF... NRC § 171.15 Annual fees: Reactor licenses and independent spent fuel storage licenses. (a) Each...

10 CFR 171.15 - Annual fees: Reactor licenses and independent spent fuel storage licenses.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Annual fees: Reactor licenses and independent spent fuel... REACTOR LICENSES AND FUEL CYCLE LICENSES AND MATERIALS LICENSES, INCLUDING HOLDERS OF CERTIFICATES OF... NRC § 171.15 Annual fees: Reactor licenses and independent spent fuel storage licenses. (a) Each...

Electrochemical cell apparatus having axially distributed entry of a fuel-spent fuel mixture transverse to the cell lengths

DOEpatents

Reichner, P.; Dollard, W.J.

1991-01-08

An electrochemical apparatus is made having a generator section containing axially elongated electrochemical cells, a fresh gaseous feed fuel inlet, a gaseous feed oxidant inlet, and at least one gaseous spent fuel exit channel, where the spent fuel exit channel passes from the generator chamber to combine with the fresh feed fuel inlet at a mixing apparatus, reformable fuel mixture channel passes through the length of the generator chamber and connects with the mixing apparatus, that channel containing entry ports within the generator chamber, where the axis of the ports is transverse to the fuel electrode surfaces, where a catalytic reforming material is distributed near the reformable fuel mixture entry ports. 2 figures.

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

NASA Astrophysics Data System (ADS)

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

2013-11-01

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

Gamma and fast neutron radiation monitoring inside spent reactor fuel assemblies

NASA Astrophysics Data System (ADS)

Lakosi, L.; Tam Nguyen, C.

2007-09-01

Gamma and neutron signatures of spent reactor fuel were monitored by small-size silicon diode and track etch detectors, respectively, in a nuclear power plant (NPP). These signatures, reflecting gross gamma intensity and the 242,244Cm content, contain information on the burn-up (BU) and cooling time (CT) of the fuel. The small size of the detectors allows close access to inside parts of the assemblies out of reach of other methods. A commercial Si diode was encapsulated in a cylindrical steel case and was used for gross γ monitoring. CR-39 detectors were used for neutron measurements. Irradiation exposures at the NPP were implemented in the central dosimetric channel of spent fuel assemblies (SFAs) stored in borated water. Gross γ and neutron axial profiles were taken up by scanning with the aid of a long steel guide tube, lowered down to the spent fuel pond by crane and fitted to the headpiece of the fuel assemblies. Gamma measurements were performed using a long cable introduced in this tube, with the Si diode at the end. A long steel wire was also led through the guide tube, to which a chain of 15 sample holder capsules was attached, each containing a track detector. Gamma dose rates of 0.1-10 kGy h -1, while neutron fluxes in a range of (0.25-26) 10 4 cm -2 s -1 were recorded. The results are in good correlation with those of a calculation for spent fuel neutron yield.

Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

NASA Astrophysics Data System (ADS)

Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

2005-02-01

A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

Determination of plutonium in spent nuclear fuel using high resolution X-ray

DOE PAGES

McIntosh, Kathryn G.; Reilly, Sean D.; Havrilla, George J.

2015-05-30

Characterization of Pu is an essential aspect of safeguards operations at nuclear fuel reprocessing facilities. A novel analysis technique called hiRX (high resolution X-ray) has been developed for the direct measurement of Pu in spent nuclear fuel dissolver solutions. hiRX is based on monochromatic wavelength dispersive X-ray fluorescence (MWDXRF), which provides enhanced sensitivity and specificity compared with conventional XRF techniques. A breadboard setup of the hiRX instrument was calibrated using spiked surrogate spent fuel (SSF) standards prepared as dried residues. Samples of actual spent fuel were utilized to evaluate the performance of the hiRX. The direct detection of just 39more » ng of Pu is demonstrated. Initial quantitative results, with error of 4–27% and precision of 2% relative standard deviation (RSD), were obtained for spent fuel samples. The limit of detection for Pu (100 s) within an excitation spot of 200 μm diameter was 375 pg. This study demonstrates the potential for the hiRX technique to be utilized for the rapid, accurate, and precise determination of Pu. Moreover, the results highlight the analytical capability of hiRX for other applications requiring sensitive and selective nondestructive analyses.« less

Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carlsen, Brett; Chiguer, Mustapha; Grahn, Per

Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years untilmore » reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is

Scientists warn of 'trillion-dollar' spent-fuel risk

NASA Astrophysics Data System (ADS)

Gwynne, Peter

2016-07-01

A study by two Princeton University physicists suggests that a major fire in the spent nuclear fuel stored on the sites of US nuclear reactors could “dwarf the horrific consequences of the Fukushima accident”.

Suggestion on the safety classification of spent fuel dry storage in China’s pressurized water reactor nuclear power plant

NASA Astrophysics Data System (ADS)

Liu, Ting; Qu, Yunhuan; Meng, De; Zhang, Qiaoer; Lu, Xinhua

2018-01-01

China’s spent fuel storage in the pressurized water reactors(PWR) is stored with wet storage way. With the rapid development of nuclear power industry, China’s NPPs(NPPs) will not be able to meet the problem of the production of spent fuel. Currently the world’s major nuclear power countries use dry storage as a way of spent fuel storage, so in recent years, China study on additional spent fuel dry storage system mainly. Part of the PWR NPP is ready to apply for additional spent fuel dry storage system. It also need to safety classificate to spent fuel dry storage facilities in PWR, but there is no standard for safety classification of spent fuel dry storage facilities in China. Because the storage facilities of the spent fuel dry storage are not part of the NPP, the classification standard of China’s NPPs is not applicable. This paper proposes the safety classification suggestion of the spent fuel dry storage for China’s PWR NPP, through to the study on China’s safety classification principles of PWR NPP in “Classification for the items of pressurized water reactor nuclear power plants (GB/T 17569-2013)”, and safety classification about spent fuel dry storage system in NUREG/CR - 6407 in the United States.

Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

NASA Astrophysics Data System (ADS)

Harkness, Ira; Zhu, Ting; Liang, Yinong; Rauch, Eric; Enqvist, Andreas; Jordan, Kelly A.

2018-01-01

Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on "fingerprinting" casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17ˣ17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

Method for storing spent nuclear fuel in repositories

DOEpatents

Schweitzer, Donald G.; Sastre, Cesar; Winsche, Warren

1981-01-01

A method for storing radioactive spent fuel in repositories containing sulfur as the storage medium is disclosed. Sulfur is non-corrosive and not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

Method for storing spent nuclear fuel in repositories

DOEpatents

Schweitzer, D.G.; Sastre, C.; Winsche, W.

A method for storing radioactive spent fuel in repositories containing sulfur as the storage medium is disclosed. Sulfur is non-corrosive and not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

Kinetic Monte Carlo Simulations and Molecular Conductance Measurements of the Bacterial Decaheme Cytochrome MtrF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Byun, H. S.; Pirbadian, S.; Nakano, Aiichiro

2014-09-05

Microorganisms overcome the considerable hurdle of respiring extracellular solid substrates by deploying large multiheme cytochrome complexes that form 20 nanometer conduits to traffic electrons through the periplasm and across the cellular outer membrane. Here we report the first kinetic Monte Carlo simulations and single-molecule scanning tunneling microscopy (STM) measurements of the Shewanella oneidensis MR-1 outer membrane decaheme cytochrome MtrF, which can perform the final electron transfer step from cells to minerals and microbial fuel cell anodes. We find that the calculated electron transport rate through MtrF is consistent with previously reported in vitro measurements of the Shewanella Mtr complex, asmore » well as in vivo respiration rates on electrode surfaces assuming a reasonable (experimentally verified) coverage of cytochromes on the cell surface. The simulations also reveal a rich phase diagram in the overall electron occupation density of the hemes as a function of electron injection and ejection rates. Single molecule tunneling spectroscopy confirms MtrF's ability to mediate electron transport between an STM tip and an underlying Au(111) surface, but at rates higher than expected from previously calculated heme-heme electron transfer rates for solvated molecules.« less

Instant release fraction corrosion studies of commercial UO2 BWR spent nuclear fuel

NASA Astrophysics Data System (ADS)

Martínez-Torrents, Albert; Serrano-Purroy, Daniel; Sureda, Rosa; Casas, Ignasi; de Pablo, Joan

2017-05-01

The instant release fraction of a spent nuclear fuel is a matter of concern in the performance assessment of a deep geological repository since it increases the radiological risk. Corrosion studies of two different spent nuclear fuels were performed using bicarbonate water under oxidizing conditions to study their instant release fraction. From each fuel, cladded segments and powder samples obtained at different radial positions were used. The results were normalised using the specific surface area to permit a comparison between fuels and samples. Different radionuclide dissolution patterns were studied in terms of water contact availability and radial distribution in the spent nuclear fuel. The relationship between the results of this work and morphological parameters like the grain size or irradiation parameters such as the burn-up or the linear power density was studied in order to increase the understanding of the instant release fraction formation.

76 FR 2277 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-13

... Fuel Storage Casks: NUHOMS[supreg] HD System Revision 1 AGENCY: Nuclear Regulatory Commission. ACTION... amend its spent fuel storage cask regulations by revising the Transnuclear, Inc. (TN) NUHOMS[supreg] HD System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1 to...

Electrochemical cell apparatus having axially distributed entry of a fuel-spent fuel mixture transverse to the cell lengths

DOEpatents

Reichner, Philip; Dollard, Walter J.

1991-01-01

An electrochemical apparatus (10) is made having a generator section (22) containing axially elongated electrochemical cells (16), a fresh gaseous feed fuel inlet (28), a gaseous feed oxidant inlet (30), and at least one gaseous spent fuel exit channel (46), where the spent fuel exit channel (46) passes from the generator chamber (22) to combine with the fresh feed fuel inlet (28) at a mixing apparatus (50), reformable fuel mixture channel (52) passes through the length of the generator chamber (22) and connects with the mixing apparatus (50), that channel containing entry ports (54) within the generator chamber (22), where the axis of the ports is transverse to the fuel electrode surfaces (18), where a catalytic reforming material is distributed near the reformable fuel mixture entry ports (54).

Dry-vault storage of spent fuel at the CASCAD facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baillif, L.; Guay, M.

A new modular dry storage vault concept using vertical metallic wells cooled by natural convection has been developed by the Commissariat a l'Energie Atomique and Societe Generale pour les Techniques Nouvelles to accommodate special fuels for high-level wastes. Basic specifications and design criteria have been followed to guarantee a double containment system and cooling to maintain the fuel below an acceptable temperature. The double containment is provided by two static barriers: At the reactor, fuels are placed in containers playing the role of the first barrier; the storage wells constitute the second barrier. Spent fuel placed in wells is cooledmore » by natural convection: a boundary layer is created along the outer side of the well. The heated air rises along the well leading to a thermosiphon flow that extracts the heat released. For heat transfer, studies, computations, and experimental tests have been carried out to calculate and determine the temperature of the containers and the fuel rod temperatures in various situations. The CASCAD vault storage can be applied to light water reactor (LWR) fuels without any difficulties if two requirements are satisfied: (1) Spend fuels have to be inserted in tight canisters. (2) Spent fuels have to be received only after a minimum decay time of 5 yr.« less

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

NASA Astrophysics Data System (ADS)

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

2017-01-01

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ∼ 18 σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Potential detector technologies and geometries are discussed.

ETR AND MTR COMPLEXES IN CONTEXT. CAMERA FACING NORTHERLY. FROM ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR AND MTR COMPLEXES IN CONTEXT. CAMERA FACING NORTHERLY. FROM BOTTOM TO TOP: ETR COOLING TOWER, ELECTRICAL BUILDING AND LOW-BAY SECTION OF ETR BUILDING, HEAT EXCHANGER BUILDING (WITH U SHAPED YARD), COMPRESSOR BUILDING. MTR REACTOR SERVICES BUILDING IS ATTACHED TO SOUTH WALL OF MTR. WING A IS ATTACHED TO BALCONY FLOOR OF MTR. NEAR UPPER RIGHT CORNER OF VIEW IS MTR PROCESS WATER BUILDING. WING B IS AT FAR WEST END OF COMPLEX. NEAR MAIN GATE IS GAMMA FACILITY, WITH "COLD" BUILDINGS BEYOND: RAW WATER STORAGE TANKS, STEAM PLANT, MTR COOLING TOWER PUMP HOUSE AND COOLING TOWER. INL NEGATIVE NO. 56-4101. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Accelerator-driven transmutation of spent fuel elements

DOEpatents

Venneri, Francesco; Williamson, Mark A.; Li, Ning

2002-01-01

An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

Direct disposal of spent fuel: developing solutions tailored to Japan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kawamura, Hideki; McKinley, Ian G

2013-07-01

With the past Government policy of 100% reprocessing in Japan now open to discussion, options for direct disposal of spent fuel (SF) are now being considered in Japan. The need to move rapidly ahead in developing spent fuel management concepts is closely related to the ongoing debate on the future of nuclear power in Japan and the desire to understand the true costs of the entire life cycle of different options. Different scenarios for future nuclear power - and associated decisions on extent of reprocessing - will give rise to quite different inventories of SF with different disposal challenges. Althoughmore » much work has been carried out spent fuel disposal within other national programmes, the potential for mining the international knowledge base is limited by the boundary conditions for disposal in Japan. Indeed, with a volunteer approach to siting, no major salt deposits and few undisturbed sediments, high tectonic activity, relatively corrosive groundwater and no deserts, it is evident that a tailored solution is needed. Nevertheless, valuable lessons can be learned from projects carried out worldwide, if focus is placed on basic principles rather than implementation details. (authors)« less

The use of experimental data in an MTR-type nuclear reactor safety analysis

NASA Astrophysics Data System (ADS)

Day, Simon E.

Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.

2013-07-01

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. Themore » initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry

Integrated Decision-Making Tool to Develop Spent Fuel Strategies for Research Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beatty, Randy L; Harrison, Thomas J

IAEA Member States operating or having previously operated a Research Reactor are responsible for the safe and sustainable management and disposal of associated radioactive waste, including research reactor spent nuclear fuel (RRSNF). This includes the safe disposal of RRSNF or the corresponding equivalent waste returned after spent fuel reprocessing. One key challenge to developing general recommendations lies in the diversity of spent fuel types, locations and national/regional circumstances rather than mass or volume alone. This is especially true given that RRSNF inventories are relatively small, and research reactors are rarely operated at a high power level or duration typical ofmore » commercial power plants. Presently, many countries lack an effective long-term policy for managing RRSNF. This paper presents results of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) #T33001 on Options and Technologies for Managing the Back End of the Research Reactor Nuclear Fuel Cycle which includes an Integrated Decision Making Tool called BRIDE (Back-end Research reactor Integrated Decision Evaluation). This is a multi-attribute decision-making tool that combines the Total Estimated Cost of each life-cycle scenario with Non-economic factors such as public acceptance, technical maturity etc and ranks optional back-end scenarios specific to member states situations in order to develop a specific member state strategic plan with a preferred or recommended option for managing spent fuel from Research Reactors.« less

77 FR 28406 - Spent Fuel Transportation Risk Assessment

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-14

... Regulations (10 CFR) part 71, ``Packaging and Transportation of Radioactive Waste,'' dated January 26, 2004) for the packaging and transport of spent nuclear fuel (and other large quantities of radioactive... NUREG- 0170, ``Final Environmental Statement on the Transportation of Radioactive Material by Air and...

Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

DOE PAGES

Hu, Jianwei; Gauld, Ian C.

2014-12-01

The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried outmore » to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.« less

Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Gauld, Ian C.

The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried outmore » to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Christian, Jerry D.; Sterbentz, James W.

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

Comparison of selected foreign plans and practices for spent fuel and high-level waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schneider, K.J.; Mitchell, S.J.; Lakey, L.T.

1990-04-01

This report describes the major parameters for management of spent nuclear fuel and high-level radioactive wastes in selected foreign countries as of December 1989 and compares them with those in the United States. The foreign countries included in this study are Belgium, Canada, France, the Federal Republic of Germany, Japan, Sweden, Switzerland, and the United Kingdom. All the countries are planning for disposal of spent fuel and/or high-level wastes in deep geologic repositories. Most countries (except Canada and Sweden) plan to reprocess their spent fuel and vitrify the resultant high-level liquid wastes; in comparison, the US plans direct disposal ofmore » spent fuel. The US is planning to use a container for spent fuel as the primary engineered barrier. The US has the most developed repository concept and has one of the earliest scheduled repository startup dates. The repository environment presently being considered in the US is unique, being located in tuff above the water table. The US also has the most prescriptive regulations and performance requirements for the repository system and its components. 135 refs., 8 tabs.« less

CONTROL CONSOLE FOR MTR FISSION PRODUCT MONITOR, USED TO DETECT ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

CONTROL CONSOLE FOR MTR FISSION PRODUCT MONITOR, USED TO DETECT BREAKS IN CLADDING OF FUEL ELEMENTS. COUNT-RATE METER IN TOP PANEL INDICATES AMOUNT OF RADIOACTIVITY. LOWER PANELS SUPPLY POWER AND AMPLIFICATION OF SIGNALS GENERATED BY SCINTILLATION COUNTER/PHOTOMULTIPLIER TUBE COMBINATION IN RESPONSE TO RADIOACTIVITY IN A SAMPLE OF THE COOLING WATER. INL NEGATIVE NO. 56-771. Jack L. Anderson, Photographer, 3/15/1956. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

10 CFR 51.23 - Temporary storage of spent fuel after cessation of reactor operation-generic determination of no...

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Temporary storage of spent fuel after cessation of reactor... Procedures § 51.23 Temporary storage of spent fuel after cessation of reactor operation—generic determination... necessary, spent fuel generated in any reactor can be stored safely and without significant environmental...

10 CFR 51.23 - Temporary storage of spent fuel after cessation of reactor operation-generic determination of no...

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Temporary storage of spent fuel after cessation of reactor... Procedures § 51.23 Temporary storage of spent fuel after cessation of reactor operation—generic determination... necessary, spent fuel generated in any reactor can be stored safely and without significant environmental...

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

DOE PAGES

Poulson, Daniel Cris; Durham, J. Matthew; Guardincerri, Elena; ...

2016-10-22

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This article describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casksmore » is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ~18σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Finally, we discuss potential detector technologies and geometries.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poulson, Daniel Cris; Durham, J. Matthew; Guardincerri, Elena

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This article describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casksmore » is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ~18σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Finally, we discuss potential detector technologies and geometries.« less

AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorgemans, J.; Mulhollem, L.; Glavin, J.

2012-07-01

The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first levelmore » of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all

Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

DOE Office of Scientific and Technical Information (OSTI.GOV)

K. J. Allen; I. Bolshinsky; L. L. Biro

2010-07-01

Romania safely air shipped 23.7 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel from the VVR S research reactor at Magurele, Romania, to the Russian Federation in June 2009. This was the world’s first air shipment of spent nuclear fuel transported in a Type B(U) cask under existing international laws without special exceptions for the air transport licenses. This shipment was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in cooperation with the Romania National Commission for Nuclear Activities Control (CNCAN), the Horiamore » Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), and the Russian Federation State Corporation Rosatom. The shipment was transported by truck to and from the respective commercial airports in Romania and the Russian Federation and stored at a secure nuclear facility in Russia where it will be converted into low enriched uranium. With this shipment, Romania became the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the work, equipment, and approvals that were required to complete this spent fuel air shipment.« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-03

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-155; 72-43 and NRC-2013-0218] Entergy Nuclear Operations, Inc.; Big Rock Point; Independent Spent Fuel Storage Installation AGENCY: Nuclear Regulatory... the Big Rock Point (BRP) Independent Spent Fuel Storage Installation (ISFSI). ADDRESSES: Please refer...

Separator assembly for use in spent nuclear fuel shipping cask

DOEpatents

Bucholz, James A.

1983-01-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dilger, Fred; Halstead, Robert J.; Ballard, James D.

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-25

... Prairie Island Nuclear Generating Plant Independent Spent Fuel Storage Installation AGENCY: Nuclear... INFORMATION CONTACT: Pamela Longmire, Ph.D., Project Manager, Licensing Branch, Division of Spent Fuel Storage... February 29, 2012 (ADAMS Accession number ML12065A073), by Prairie Island Nuclear Generating Plant (PINGP...

Depleted uranium startup of spent-fuel treatment operations at ANL-West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goff, K.M.; Mariani, R.D.; Bonomo, N.L.

1995-12-31

At Argonne National Laboratory-West (ANL-West) there are several thousand kilograms of Experimental Breeder Reactor II (EBR-II) spent nuclear fuel. This fuel will be treated using an electrometallurgical process in the fuel conditioning facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. The process equipment is undergoing testing with depleted uranium in preparation for irradiated fuel operations during the summer of 1995.

THE ECONOMICS OF REPROCESSING vs DIRECT DISPOSAL OF SPENT NUCLEAR FUEL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Matthew Bunn; Steve Fetter; John P. Holdren

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2011-07-18

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

Watkins: Store spent fuel at federal sites

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tompkins, B.

If the Department of Energy under the Clinton Administration decides to follow the strategy of departing Energy Secretary James Watkins, spent nuclear fuel (SNF) from utilities would be stored at federal sites starting in 1998. In what amounts to an admission that a monitored retrievable storage (MRS) facility will not be ready to accept SNF by 1998, Watkins proposed in a December 17 letter to Senator J. Bennett Johnston that current work on an MRS be terminated, and that efforts be redirected toward standardized spent-fuel container design. According to the DOE's Samantha Williams, however, the Waste Negotiator's Office will continuemore » its activities, and the new strategy is intended to supplement the ongoing MRS work. She also said that DOE Undersecretary Hugo Pomrehn confirmed this in a statement to the Nuclear Waste TEchnical Review Board during its meeting on January 5. The motivation for the new strategic actions is in part related to the passage of the National Energy Policy Act of 1992, which envisions the continued use of nuclear power in the overall US energy plan. Watkins points out, however, that in order for nuclear power to remain a viable alternative, the management and disposal of spent commercial nuclear fuel is essential to avoid premature and unwarranted shutdowns of operating nuclear plants, to permit renewals of existing plant licenses to provide life extensions from 40 to 60 years, and to enable new orders of advanced-design nuclear plants. In addition, the new strategy is required, according to Watkins, because the Office of the Nuclear Waste Negotiator has been unsuccessful in identifying a voluntary host site for an MRS after two years of efforts to that end. David Leroy, the Negotiator, disagrees with this statement, and has noted that it still may be possible to present a candidate host site for the MRS to Congress by June 1993.« less

Method for calculating the duration of vacuum drying of a metal-concrete container for spent nuclear fuel

NASA Astrophysics Data System (ADS)

Karyakin, Yu. E.; Nekhozhin, M. A.; Pletnev, A. A.

2013-07-01

A method for calculating the quantity of moisture in a metal-concrete container in the process of its charging with spent nuclear fuel is proposed. A computing method and results obtained by it for conservative estimation of the time of vacuum drying of a container charged with spent nuclear fuel by technologies with quantization and without quantization of the lower fuel element cluster are presented. It has been shown that the absence of quantization in loading spent fuel increases several times the time of vacuum drying of the metal-concrete container.

COBRA-SFS thermal-hydraulic analysis code for spent fuel storage and transportation casks: Models and methods

DOE PAGES

Michener, Thomas E.; Rector, David R.; Cuta, Judith M.

2017-09-01

COBRA-SFS, a thermal-hydraulics code developed for steady-state and transient analysis of multi-assembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent fuel package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is themore » capability for detailed thermal radiation modeling within the fuel rod array.« less

COBRA-SFS thermal-hydraulic analysis code for spent fuel storage and transportation casks: Models and methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michener, Thomas E.; Rector, David R.; Cuta, Judith M.

COBRA-SFS, a thermal-hydraulics code developed for steady-state and transient analysis of multi-assembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent fuel package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is themore » capability for detailed thermal radiation modeling within the fuel rod array.« less

Study of Compton suppression for use in spent nuclear fuel assay

NASA Astrophysics Data System (ADS)

Bender, Sarah

The focus of this study has been to assess Compton suppressed gamma-ray detection systems for the multivariate analysis of spent nuclear fuel. This objective has been achieved using direct measurement of samples of irradiated fuel elements in two geometrical configurations with Compton suppression systems. In order to address the objective to quantify the number of additionally resolvable photopeaks, direct Compton suppressed spectroscopic measurements of spent nuclear fuel in two configurations were performed: as intact fuel elements and as dissolved feed solutions. These measurements directly assessed and quantified the differences in measured gamma-ray spectrum from the application of Compton suppression. Several irradiated fuel elements of varying cooling time from the Penn State Breazeale Reactor spent fuel inventory were measured using three Compton suppression systems that utilized different primary detectors: HPGe, LaBr3, and NaI(Tl). The application of Compton suppression using a LaBr3 primary detector to the measurement of the current core fuel element, which presented the highest count rate, allowed four additional spectral features to be resolved. In comparison, the HPGe-CSS was able to resolve eight additional photopeaks as compared to the standalone HPGe measurement. Measurements with the NaI(Tl) primary detector were unable to resolve any additional peaks, due to its relatively low resolution. Samples of Approved Test Material (ATM) commercial fuel elements were obtained from Pacific Northwest National Laboratory. The samples had been processed using the beginning stages of the PUREX method and represented the unseparated feed solution from a reprocessing facility. Compton suppressed measurements of the ATM fuel samples were recorded inside the guard detector annulus, to simulate the siphoning of small quantities from the main process stream for long dwell measurement periods. Photopeak losses were observed in the measurements of the dissolved ATM

Dry halide method for separating the components of spent nuclear fuels

DOEpatents

Christian, J.D.; Thomas, T.R.; Kessinger, G.F.

1998-06-30

The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200 C to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400 C; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164 to 2 C; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic. 3 figs.

Dry halide method for separating the components of spent nuclear fuels

DOEpatents

Christian, Jerry Dale; Thomas, Thomas Russell; Kessinger, Glen F.

1998-01-01

The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission- and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200.degree. C. to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400.degree. C.; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164.degree. C. to 2.degree. C.; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic.

Structural Health Monitoring of Nuclear Spent Fuel Storage Facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yu, Lingyu

Interim storage of spent nuclear fuel from reactor sites has gained additional importance and urgency for resolving waste-management-related technical issues. To ensure that nuclear power remains clean energy, monitoring has been identified by DOE as a high priority cross-cutting need, necessary to determine and predict the degradation state of the systems, structures, and components (SSCs) important to safety (ITS). Therefore, nondestructive structural condition monitoring becomes a need to be installed on existing or to be integrated into future storage system to quantify the state of health or to guarantee the safe operation of nuclear power plants (NPPs) during their extendedmore » life span. In this project, the lead university and the collaborating national laboratory teamed to develop a nuclear structural health monitoring (n-SHM) system based on in-situ piezoelectric sensing technologies that can monitor structural degradation and aging for nuclear spent fuel DCSS and similar structures. We also aimed to identify and quantify possible influences of nuclear spent fuel environment (temperature and radiation) to the piezoelectric sensor system and come up with adequate solutions and guidelines therefore. We have therefore developed analytical model for piezoelectric based n-SHM methods, with considerations of temperature and irradiation influence on the model of sensing and algorithms in acoustic emission (AE), guided ultrasonic waves (GUW), and electromechanical impedance spectroscopy (EMIS). On the other side, experimentally the temperature and irradiation influence on the piezoelectric sensors and sensing capabilities were investigated. Both short-term and long-term irradiation investigation with our collaborating national laboratory were performed. Moreover, we developed multi-modal sensing, validated in laboratory setup, and conducted the testing on the We performed multi-modal sensing development, verification and validation tests on very complex

EURATOM safeguards efforts in the development of spent fuel verification methods by non-destructive assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Matloch, L.; Vaccaro, S.; Couland, M.

The back end of the nuclear fuel cycle continues to develop. The European Commission, particularly the Nuclear Safeguards Directorate of the Directorate General for Energy, implements Euratom safeguards and needs to adapt to this situation. The verification methods for spent nuclear fuel, which EURATOM inspectors can use, require continuous improvement. Whereas the Euratom on-site laboratories provide accurate verification results for fuel undergoing reprocessing, the situation is different for spent fuel which is destined for final storage. In particular, new needs arise from the increasing number of cask loadings for interim dry storage and the advanced plans for the construction ofmore » encapsulation plants and geological repositories. Various scenarios present verification challenges. In this context, EURATOM Safeguards, often in cooperation with other stakeholders, is committed to further improvement of NDA methods for spent fuel verification. In this effort EURATOM plays various roles, ranging from definition of inspection needs to direct participation in development of measurement systems, including support of research in the framework of international agreements and via the EC Support Program to the IAEA. This paper presents recent progress in selected NDA methods. These methods have been conceived to satisfy different spent fuel verification needs, ranging from attribute testing to pin-level partial defect verification. (authors)« less

76 FR 17037 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-28

...-0007] RIN 3150-AI90 List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition AGENCY... or the Commission) is proposing to amend its spent fuel storage cask regulations to add the HI-STORM...: June 13, 2011. SAR Submitted by: Holtec International, Inc. SAR Title: Safety Analysis Report on the HI...

MTR, SOUTH FACE OF REACTOR. SPECIAL SUPPLEMENTAL SHIELDING WAS REQUIRED ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR, SOUTH FACE OF REACTOR. SPECIAL SUPPLEMENTAL SHIELDING WAS REQUIRED OUTSIDE OF MTR FOR EXPERIMENTS. THE AIRCRAFT NUCLEAR PROPULSION PROJECT DOMINATED THE USE OF THIS PART OF THE MTR. INL NEGATIVE NO. 7225. Unknown Photographer, 11/28/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

The MATROSHKA experiment: results and comparison from extravehicular activity (MTR-1) and intravehicular activity (MTR-2A/2B) exposure.

PubMed

Berger, Thomas; Bilski, Paweł; Hajek, Michael; Puchalska, Monika; Reitz, Günther

2013-12-01

Astronauts working and living in space are exposed to considerably higher doses and different qualities of ionizing radiation than people on Earth. The multilateral MATROSHKA (MTR) experiment, coordinated by the German Aerospace Center, represents the most comprehensive effort to date in radiation protection dosimetry in space using an anthropomorphic upper-torso phantom used for radiotherapy treatment planning. The anthropomorphic upper-torso phantom maps the radiation distribution as a simulated human body installed outside (MTR-1) and inside different compartments (MTR-2A: Pirs; MTR-2B: Zvezda) of the Russian Segment of the International Space Station. Thermoluminescence dosimeters arranged in a 2.54 cm orthogonal grid, at the site of vital organs and on the surface of the phantom allow for visualization of the absorbed dose distribution with superior spatial resolution. These results should help improve the estimation of radiation risks for long-term human space exploration and support benchmarking of radiation transport codes.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-05

... processing is a chemical separations process that involves dissolving spent fuel in nitric acid and... Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact... chemical properties, and radionuclide inventory. The fuel groups and the seven technologies that could be...

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

NASA Astrophysics Data System (ADS)

Casella, Andrew M.

2007-09-01

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

Technology, safety and costs of decommissioning reference independent spent fuel storage installations. [Contains glossary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ludwick, J D; Moore, E B

1984-01-01

Safety and cost information is developed for the conceptual decommissioning of five different types of reference independent spent fuel storage installations (ISFSIs), each of which is being given consideration for interim storage of spent nuclear fuel in the United States. These include one water basin-type ISFSI (wet) and four dry ISFSIs (drywell, silo, vault, and cask). The reference ISFSIs include all component parts necessary for the receipt, handling and storage of spent fuel in a safe and efficient manner. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, and potential radiation doses tomore » the public. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment followed by long-term surveillance).« less

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

DOE PAGES

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek; ...

2017-03-27

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

Characteristics and Dose Levels for Spent Reactor Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coates, Cameron W

2007-01-01

Current guidance considers highly radioactive special nuclear materials to be those materials that, unshielded, emit a radiation dose [rate] measured at 1 m which exceeds 100 rem/h. Smaller, less massive fuel assemblies from research reactors can present a challenge from the point of view of self protection because of their size (lower dose, easier to handle) and the desirability of higher enrichments; however, a follow-on study to cross-compare dose trends of research reactors and power reactors was deemed useful to confirm/verify these trends. This paper summarizes the characteristics and dose levels of spent reactor fuels for both research reactors andmore » power reactors and extends previous studies aimed at quantifying expected dose rates from research reactor fuels worldwide.« less

Radio-toxicity of spent fuel of the advanced heavy water reactor.

PubMed

Anand, S; Singh, K D S; Sharma, V K

2010-01-01

The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR.

Fate of Noble Metals during the Pyroprocessing of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

B.R. Westphal; D. Vaden; S.X. Li

During the pyroprocessing of spent nuclear fuel by electrochemical techniques, fission products are separated as the fuel is oxidized at the anode and refined uranium is deposited at the cathode. Those fission products that are oxidized into the molten salt electrolyte are considered active metals while those that do not react are considered noble metals. The primary noble metals encountered during pyroprocessing are molybdenum, zirconium, ruthenium, rhodium, palladium, and technetium. Pyroprocessing of spent fuel to date has involved two distinctly different electrorefiner designs, in particular the anode to cathode configuration. For one electrorefiner, the anode and cathode collector are horizontallymore » displaced such that uranium is transported across the electrolyte medium. As expected, the noble metal removal from the uranium during refining is very high, typically in excess of 99%. For the other electrorefiner, the anode and cathode collector are vertically collocated to maximize uranium throughput. This arrangement results in significantly less noble metals removal from the uranium during refining, typically no better than 20%. In addition to electrorefiner design, operating parameters can also influence the retention of noble metals, albeit at the cost of uranium recovery. Experiments performed to date have shown that as much as 100% of the noble metals can be retained by the cladding hulls while affecting the uranium recovery by only 6%. However, it is likely that commercial pyroprocessing of spent fuel will require the uranium recovery to be much closer to 100%. The above mentioned design and operational issues will likely be driven by the effects of noble metal contamination on fuel fabrication and performance. These effects will be presented in terms of thermal properties (expansion, conductivity, and fusion) and radioactivity considerations. Ultimately, the incorporation of minor amounts of noble metals from pyroprocessing into fast reactor

Structural basis for MTR4-ZCCHC8 interactions that stimulate the MTR4 helicase in the nuclear exosome-targeting complex.

PubMed

Puno, M Rhyan; Lima, Christopher D

2018-06-12

The nuclear exosome-targeting (NEXT) complex functions as an RNA exosome cofactor and is involved in surveillance and turnover of aberrant transcripts and noncoding RNAs. NEXT is a ternary complex composed of the RNA-binding protein RBM7, the scaffold zinc-knuckle protein ZCCHC8, and the helicase MTR4. While RNA interactions with RBM7 are known, it remains unclear how NEXT subunits collaborate to recognize and prepare substrates for degradation. Here, we show that MTR4 helicase activity is enhanced when associated with RBM7 and ZCCHC8. While uridine-rich substrates interact with RBM7 and are preferred, optimal activity is observed when substrates include a polyadenylated 3' end. We identify a bipartite interaction of ZCCHC8 with MTR4 and uncover a role for the conserved C-terminal domain of ZCCHC8 in stimulating MTR4 helicase and ATPase activities. A crystal structure reveals that the ZCCHC8 C-terminal domain binds the helicase core in a manner that is distinct from that observed for Saccharomyces cerevisiae exosome cofactors Trf4p and Air2p. Our results are consistent with a model whereby effective targeting of substrates by NEXT entails recognition of elements within the substrate and activation of MTR4 helicase activity. Copyright © 2018 the Author(s). Published by PNAS.

75 FR 27401 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1; Correction

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-17

... Storage Casks: NUHOMS[reg] HD System Revision 1; Correction AGENCY: Nuclear Regulatory Commission. ACTION... HD spent fuel storage cask system. This action is necessary to correctly specify the effective date... on May 6, 2010 (75 FR 24786), that amends the regulations that govern storage of spent nuclear fuel...

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

DOEpatents

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

1982-01-19

A method and apparatus for separating and reprocessing spent nuclear fuels includes a separation vessel housing a molten metal solvent in a reaction region, a reflux region positioned above and adjacent to the reaction region, and a porous filter member defining the bottom of the separation vessel in a supporting relationship with the metal solvent. Spent fuels are added to the metal solvent. A nonoxidizing nitrogen-containing gas is introduced into the separation vessel, forming solid actinide nitrides in the metal solvent from actinide fuels, while leaving other fission products in solution. A pressure of about 1.1 to 1.2 atm is applied in the reflux region, forcing the molten metal solvent and soluble fission products out of the vessel, while leaving the solid actinide nitrides in the separation vessel.

Nuclear Forensics Attributing the Source of Spent Fuel Used in an RDD Event

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scott, Mark Robert

2005-05-01

An RDD attack against the U.S. is something America needs to prepare against. If such an event occurs the ability to quickly identify the source of the radiological material used in an RDD would aid investigators in identifying the perpetrators. Spent fuel is one of the most dangerous possible radiological sources for an RDD. In this work, a forensics methodology was developed and implemented to attribute spent fuel to a source reactor. The specific attributes determined are the spent fuel burnup, age from discharge, reactor type, and initial fuel enrichment. It is shown that by analyzing the post-event material, thesemore » attributes can be determined with enough accuracy to be useful for investigators. The burnup can be found within a 5% accuracy, enrichment with a 2% accuracy, and age with a 10% accuracy. Reactor type can be determined if specific nuclides are measured. The methodology developed was implemented into a code call NEMASYS. NEMASYS is easy to use and it takes a minimum amount of time to learn its basic functions. It will process data within a few minutes and provide detailed information about the results and conclusions.« less

77 FR 26050 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-02

... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... of pressurized water reactor spent nuclear fuel (SNF) in transportation packages and storage casks... for the licensing basis, (b) provide recommendations regarding advanced isotopic depletion and...

Tools Developed to Prepare and Stabilize Reactor Spent Fuel for Retrieval from Tile Holes - 12251

DOE Office of Scientific and Technical Information (OSTI.GOV)

Horne, Michael; Clough, Malcolm

Spent fuel from the Chalk River Laboratories (CRL) nuclear reactors is stored in the waste management areas on site. This fuel is contained within carbon steel spent fuel cans that are stored inside vertical carbon steel lined concrete pipes in the ground known as tile holes. The fuel cans have been stored in the tile holes for greater than 30 years. Some of the fuel cans have experienced corrosion which may have affected their structural integrity as well as the potential to form hydrogen gas. In addition to these potential hazards, there was a need to clean contaminated surfaces insidemore » of and around the exposed upper surface of the tile holes. As part of the site waste management remediation plan spent fuel will be retrieved from degraded tile holes, dried, and relocated to a new purpose built above ground storage facility. There have been a number of tools that are required to be developed to ensure spent fuel cans are in a safe condition prior to retrieval and re-location. A series of special purpose tools have been designed and constructed to stabilize the contents of the tile holes, to determine the integrity of the fuel containers and to decontaminate inside and around the tile holes. Described herein are the methods and types of tools used. Tools that have been presented here have been used, or will be used in the near future, in the waste management areas of the CRL Site in preparation for storage of spent fuel in a new above ground facility. The stabilization tools have been demonstrated on mock-up facilities prior to successful use in the field to remove hydrogen gas and uranium hydrides from the fuel cans. A lifting tool has been developed and used successfully in the field to confirm the integrity of the fuel cans for future relocation. A tool using a commercial dry ice blaster has been developed and is ready to start mock-up trials and is scheduled to be used in the field during the summer of 2012. (authors)« less

Measurement and analysis of gamma-rays emitted from spent nuclear fuel above 3 MeV.

PubMed

Rodriguez, Douglas C; Anderson, Elaina; Anderson, Kevin K; Campbell, Luke W; Fast, James E; Jarman, Kenneth; Kulisek, Jonathan; Orton, Christopher R; Runkle, Robert C; Stave, Sean

2013-12-01

The gamma-ray spectrum of spent nuclear fuel in the 3-6 MeV energy range is important for active interrogation since gamma rays emitted from nuclear decay are not expected to interfere with measurements in this energy region. There is, unfortunately, a dearth of empirical measurements from spent nuclear fuel in this region. This work is an initial attempt to partially fill this gap by presenting an analysis of gamma-ray spectra collected from a set of spent nuclear fuel sources using a high-purity germanium detector array. This multi-crystal array possesses a large collection volume, providing high energy resolution up to 16 MeV. The results of these measurements establish the continuum count-rate in the energy region between 3 and 6 MeV. Also assessed is the potential for peaks from passive emissions to interfere with peak measurements resulting from active interrogation delayed emissions. As one of the first documented empirical measurements of passive emissions from spent fuel for energies above 3 MeV, this work provides a foundation for active interrogation model validation and detector development. © 2013 Elsevier Ltd. All rights reserved.

77 FR 60479 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-03

... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... 3, entitled, ``Burnup Credit in the Criticality Safety Analyses of PWR [Pressurized Water Reactor... water reactor spent nuclear fuel (SNF) in transportation packages and storage casks. SFST-ISG-8...

Method For Processing Spent (Trn,Zr)N Fuel

DOEpatents

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.

78 FR 78285 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-26

...-2012-0052] RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment... document proposed to amend the NRC's spent fuel storage regulations by revising the Holtec International HI...

Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

NASA Astrophysics Data System (ADS)

Johnson, Lawrence; Ferry, Cécile; Poinssot, Christophe; Lovera, Patrick

2005-11-01

A source-term model for the short-term release of radionuclides from spent nuclear fuel (SNF) has been developed. It provides quantitative estimates of the fraction of various radionuclides that are expected to be released rapidly (the instant release fraction, or IRF) when water contacts the UO 2 or MOX fuel after container breaching in a geological repository. The estimates are based on correlation of leaching data for radionuclides with fuel burnup and fission gas release. Extrapolation of the data to higher fuel burnup values is based on examination of data on fuel restructuring, such as rim development, and on fission gas release data, which permits bounding IRF values to be estimated assuming that radionuclide releases will be less than fission gas release. The consideration of long-term solid-state changes influencing the IRF prior to canister breaching is addressed by evaluating alpha self-irradiation enhanced diffusion, which may gradually increase the accumulation of fission products at grain boundaries.

Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koji Shirai

2006-04-01

The VSC-17 Spent Nuclear Fuel Storage Cask was surveyed for degradation of the concrete shield by radiation measurement, temperature measurement, and ultrasonic testing. No general loss of shielding function was identified.

Maintaining Continuity of Knowledge of Spent Fuel Pools: Tool Survey

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benz, Jacob M.; Smartt, Heidi A.; Tanner, Jennifer E.

This report examines supplemental tools that can be used in addition to optical surveillance cameras to maintain CoK in low-to-no light conditions, and increase the efficiency and effectiveness of spent fuel CoK, including item counting and ID verification, in challenging conditions.

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

DOE PAGES

Martinik, Tomas; Henzl, Vladimir; Grape, Sophie; ...

2015-03-04

Here, previous simulation studies of Differential Die–Away (DDA) instrument’s response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument’s response to interrogation of asymmetricallymore » burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martinik, Tomas; Henzl, Vladimir; Grape, Sophie

Here, previous simulation studies of Differential Die–Away (DDA) instrument’s response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument’s response to interrogation of asymmetricallymore » burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs.« less

Preliminary developments of MTR plates with uranium nitride

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durand, J.P.; Laudamy, P.; Richter, K.

1997-08-01

In the opinion of CERCA, the total weight of Uranium per MTR plate (without changing the external dimensions) cannot be further increased using U{sub 3}Si{sub 2}. Limits have been reached on plates with a thicker meat or loaded to 6g Ut/cm{sup 3}. The use of a denser fuel like Uranium mononitride could permit an increase in these limits. A collaboration between the Institute for Transuranium Elements (ITU), Joint Research Centre of the European Commission, and CERCA has been set ut. The preliminary studies at the ITU to check compatibility between aluminium and UN proved that there are no metallurgical interactionsmore » below 500{degrees}C. Feasibility of the manufacturing, on a laboratory scale at CERCA, of depleted Uranium mononitride plates loaded to 7 g Ut/cm{sup 3} has been demonstrated. The manufacturing process, however, is only one aspect of the development of a new fuel. The experience gained in the case of U{sub 3}Si{sub 2} has shown that the development of a new fuel requires considerable time and financial investment. Such a development certainly represents an effort of about 10 years.« less

The shutdown reactor: Optimizing spent fuel storage cost

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pennington, C.W.

1995-12-31

Several studies have indicated that the most prudent way to store fuel at a shutdown reactor site safely and economically is through the use of a dry storage facility licensed under 10CFR72. While such storage is certainly safe, is it true that the dry ISFSI represents the safest and most economical approach for the utility? While no one is really able to answer that question definitely, as yet, Holtec has studied this issue for some time and believes that both an economic and safety case can be made for an optimization strategy that calls for the use of both wetmore » and dry ISFSI storage of spent fuel at some plants. For the sake of brevity, this paper summarizes some of Holtec`s findings with respect to the economics of maintaining some fuel in wet storage at a shutdown reactor. The safety issue, or more importantly the perception of safety of spent fuel in wet storage, still varies too much with the eye of the beholder, and until a more rigorous presentation of safety analyses can be made in a regulatory setting, it is not practically useful to argue about how many angels can sit on the head of a safety-related pin. Holtec is prepared to present such analyses, but this does not appear to be the proper venue. Thus, this paper simply looks at certain economic elements of a wet ISFSI at a shutdown reactor to make a prima facie case that wet storage has some attractiveness at a shutdown reactor and should not be rejected out of hand. Indeed, an optimization study at certain plants may well show the economic vitality of keeping some fuel in the pool and converting the NRC licensing coverage from 10CFR50 to 10CFR72. If the economics look attractive, then the safety issue may be confronted with a compelling interest.« less

Global Spent Fuel Logistics Systems Study (GSFLS). Volume 2. GSFLS visit findings and evaluations. Interim report

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1978-01-31

This report is a part of the interim report documentation for the Global Spent Fuel Logistics System (GSFLS) study. This report describes a global framework that evaluates spent fuel disposition requirements, influencing factors and strategies. A broad sampling of foreign governmental officials, electric utility spokesmen and nuclear power industry officials responsible for GSFLS policies, plans and programs were surveyed as to their views with respect to national and international GSFLS related considerations. The results of these GSFLS visit findings are presented herein. These findings were then evaluated in terms of technical, institutional and legal/regulatory implications. The GSFLS evaluations, in conjunctionmore » with perceived US spent fuel objectives, formed the basis for selecting a set of GSFLS strategies which are reported herein.« less

Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal

NASA Astrophysics Data System (ADS)

Herrero, J. J.; Rochman, D.; Leray, O.; Vasiliev, A.; Pecchia, M.; Ferroukhi, H.; Caruso, S.

2017-09-01

In the design of a spent nuclear fuel disposal system, one necessary condition is to show that the configuration remains subcritical at time of emplacement but also during long periods covering up to 1,000,000 years. In the context of criticality safety applying burn-up credit, k-eff eigenvalue calculations are affected by nuclear data uncertainty mainly in the burnup calculations simulating reactor operation and in the criticality calculation for the disposal canister loaded with the spent fuel assemblies. The impact of nuclear data uncertainty should be included in the k-eff value estimation to enforce safety. Estimations of the uncertainty in the discharge compositions from the CASMO5 burn-up calculation phase are employed in the final MCNP6 criticality computations for the intact canister configuration; in between, SERPENT2 is employed to get the spent fuel composition along the decay periods. In this paper, nuclear data uncertainty was propagated by Monte Carlo sampling in the burn-up, decay and criticality calculation phases and representative values for fuel operated in a Swiss PWR plant will be presented as an estimation of its impact.

3DD - Three Dimensional Disposal of Spent Nuclear Fuel - 12449

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dvorakova, Marketa; Slovak, Jiri

2012-07-01

Three dimensional disposal is being considered as a way in which to store long-term spent nuclear fuel in underground disposal facilities in the Czech Republic. This method involves a combination of the two most common internationally recognised disposal methods in order to practically apply the advantages of both whilst, at the same time, eliminating their weaknesses; the method also allows easy removal in case of potential re-use. The proposed method for the disposal of spent nuclear fuel will reduce the areal requirements of future deep geological repositories by more than 30%. It will also simplify the container handling process bymore » using gravitational forces in order to meet requirements concerning the controllability of processes and ensuring operational and nuclear safety. With regard to the issue of the efficient potential removal of waste containers, this project offers an ingenious solution which does not disrupt the overall stability of the original disposal complex. (authors)« less

Default operational intervention levels (OILs) for severe nuclear power plant or spent fuel pool emergencies.

PubMed

McKenna, T; Kutkov, V; Vilar Welter, P; Dodd, B; Buglova, E

2013-05-01

Experience and studies show that for an emergency at a nuclear power plant involving severe core damage or damage to the fuel in spent fuel pools, the following actions may need to be taken in order to prevent severe deterministic health effects and reduce stochastic health effects: (1) precautionary protective actions and other response actions for those near the facility (i.e., within the zones identified by the International Atomic Energy Agency) taken immediately upon detection of facility conditions indicating possible severe damage to the fuel in the core or in the spent fuel pool; and (2) protective actions and other response actions taken based on environmental monitoring and sampling results following a release. This paper addresses the second item by providing default operational intervention levels [OILs, which are similar to the U.S. derived response levels (DRLs)] for promptly assessing radioactive material deposition, as well as skin, food, milk and drinking water contamination, following a major release of fission products from the core or spent fuel pool of a light water reactor (LWR) or a high power channel reactor (RBMK), based on the International Atomic Energy Agency's guidance.

33 CFR 154.1041 - Specific response information to be maintained on mobile MTR facilities.

Code of Federal Regulations, 2010 CFR

2010-07-01

... be maintained on mobile MTR facilities. 154.1041 Section 154.1041 Navigation and Navigable Waters... maintained on mobile MTR facilities. (a) Each mobile MTR facility must carry the following information as... respond to a discharge from the mobile MTR facility. (3) List of the appropriate persons and agencies...

78 FR 45575 - Duke Energy Carolinas, LLC; Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-29

... will allow Oconee to effectively manage its spent fuel inventory to meet decay heat zoning requirements... thermal stresses, including potential elongation from decay heat and irradiation. In addition, the NRC...] system provides for the horizontal dry storage of canisterized spent fuel assemblies in a concrete...

Modelling the radiolytic corrosion of α-doped UO2 and spent nuclear fuel

NASA Astrophysics Data System (ADS)

Liu, Nazhen; Qin, Zack; Noël, James J.; Shoesmith, David W.

2017-10-01

A model previously developed to predict the corrosion rate of spent fuel (UO2) inside a failed waste container has been adapted to simulate the rates measured on a wide range of α-doped UO2 and spent fuel specimens. This simulation confirms the validity of the model and demonstrates that the steady-state corrosion rate is controlled by the radiolytic production of H2O2 (which has been shown to be the primary oxidant driving fuel corrosion), irrespective of the reactivity of the UO2 matrix. The model was then used to determine the consequences of corrosion inside a failed container resealed by steel corrosion products. The possible accumulation of O2, produced by H2O2 decomposition, was found to accelerate the corrosion rate in a closed system. However, the simultaneous accumulation of radiolytic H2, which is activated as a reductant on the noble metal (ε) particles in the spent fuel, rapidly overcame this acceleration leading to the eventual suppression of the corrosion rate to insignificant values. Calculations also showed that, while the radiation dose rate, the H2O2 decomposition ratio, and the surface coverage of ε particles all influenced the short term corrosion rate, the influence of the radiolytically produced H2 was the overwhelming influence in reducing the rate to negligible level (i.e., <10-20 mol m-2 s-1).

324 Building spent fuel segments pieces and fragments removal summary report

DOE Office of Scientific and Technical Information (OSTI.GOV)

SMITH, C L

2003-01-09

As part of the 324 Building Deactivation Project, all Spent Nuclear Fuel (SNF) and Special Nuclear Material were removed. The removal entailed packaging the material into a GNS-12 cask and shipping it to the Central Waste Complex (CWC).

Molten tin reprocessing of spent nuclear fuel elements. [Patent application; continuous process

DOEpatents

Heckman, R.A.

1980-12-19

A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support te liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

MTR AND ETR COMPLEXES. CAMERA FACING EASTERLY TOWARD CHEMICAL PROCESSING ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR AND ETR COMPLEXES. CAMERA FACING EASTERLY TOWARD CHEMICAL PROCESSING PLANT. MTR AND ITS ATTACHMENTS IN FOREGROUND. ETR BEYOND TO RIGHT. INL NEGATIVE NO. 56-4100. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Amplicon Sequencing Reveals Microbiological Signatures in Spent Nuclear Fuel Storage Basins

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bagwell, Christopher E.; Noble, Peter A.; Milliken, Charles E.

Water quality is an important determinant for the structural integrity of alloy cladded fuels and assemblies during long-term wet storage. Detailed characterization of a water filled storage basin for spent nuclear reactor fuel was performed following the formation and proliferation of an amorphous white flocculent. White precipitant was sampled throughout the storage basin for chemical and spectroscopic characterization, and eDNA was extracted for pyrosequencing of bacterial rRNA gene diversity. Accordingly, spectroscopic analyses indicated the precipitant to be primarily amorphous to crystalline aluminum (oxy) hydroxides with minor associated elemental components including Fe, Si, Ti, and U. High levels of dissolved carbonmore » were co-localized with the precipitant relative to bulk water. Bacterial densities were highly variable between sampling locations and with depth; cell numbers (log scale) ranged from 5.6 to 4.89 cells / mL. Bacterial diversity that was physically associated with the aluminum (oxy) hydroxide complexes exceeded an estimated 4,000 OTUs / amplicon library (3% cutoff) and the greatest percent majority of sequences were aligned to the families Burkholderiales (23%), Nitrospiraceae (23%), Hyphomicrobiaceae (17%), and Comamonadaceae (6%). We surmise that episodic changes in the physical and chemical properties of the basin contribute to the polymerization of aluminum (oxy) hydroxides, which in turn can chemisorb nutrients, carbon ligands and bacterial cells from the surrounding bulk aqueous phase. As such, these precipitants should establish favorable microhabitats for bacterial colonization and growth. Comparative analyses of 16S rRNA gene amplicon libraries across diverse environmental landscapes were performed and microbiological signatures unique to the spent nuclear fuel storage basin environment were revealed. These insights could spur the development of tractable bioindicators that are specific of and diagnostic for water quality at

MRT fuel element inspection at Dounreay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gibson, J.

1997-08-01

To ensure that their production and inspection processes are performed in an acceptable manner, ie. auditable and traceable, the MTR Fuel Element Fabrication Plant at Dounreay operates to a documented quality system. This quality system, together with the fuel element manufacturing and inspection operations, has been independently certified to ISO9002-1987, EN29002-1987 and BS5750:Pt2:1987 by Lloyd`s Register Quality Assurance Limited (LRQA). This certification also provides dual accreditation to the relevant German, Dutch and Australian certification bodies. This paper briefly describes the quality system, together with the various inspection stages involved in the manufacture of MTR fuel elements at Dounreay.

Radiation Templates of Spent Fuel in Casks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vanier, Peter

BNL and INL propose to perform a scoping study, using heavily collimated gamma and fast neutron detectors, to obtain passive radiation templates of dry storage casks containing spent fuel. The goal is to demonstrate sufficient spatial resolution and sensitivity to detect a missing fuel assembly. Such measurements, combined with detailed modeling and decay corrections should provide confidence that the cask contents have not been altered, despite loss of continuity of knowledge (CoK). The concept relies on the leakage of high energy gammas and neutrons through the shielding of the casks. Tests will emphasize organic scintillators with pulse shape discrimination, butmore » baseline comparisons will be made to high purity germanium (HPGe) and collimated moderated 3He detectors deployed in the same locations. Commercial off-the-shelf (COTS) detectors and data acquisition electronics will be used with custom-built collimators and shielding.« less

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

DOEpatents

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

1982-01-19

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

10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation. The...

10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation. The...

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

PubMed

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

2005-01-01

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

ENGINEERING TEST REACTOR, TRA642. CONTEXTUAL VIEW ORIENTATING ETR TO MTR. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ENGINEERING TEST REACTOR, TRA-642. CONTEXTUAL VIEW ORIENTATING ETR TO MTR. CAMERA IS ON ROOF OF MTR BUILDING AND FACES DUE SOUTH. MTR SERVICE BUILDING, TRA-635, IN LOWER RIGHT CORNER. STEEL FRAMES SHOW BUILDINGS TO BE ATTACHED TO ETR BUILDING. HIGH-BAY SECTION IN CENTER IS REACTOR BUILDING. TWO-STORY CONTROL ROOM AND OFFICE BUILDING, TRA-647, IS BETWEEN IT AND MTR SERVICE BUILDING. STRUCTURE TO THE LEFT (WITH NO FRAMING YET) IS COMPRESSOR BUILDING, TRA-643, AND BEYOND IT WILL BE HEAT EXCHANGER BUILDING, TRA-644, GREAT SOUTHERN BUTTE ON HORIZON. INL NEGATIVE NO. 56-2382. Jack L. Anderson, Photographer, 6/10/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Modelling of radiation field around spent fuel container.

PubMed

Kryuchkov, E F; Opalovsky, V A; Tikhomirov, G V

2005-01-01

Operation of nuclear reactors leads to the production of spent nuclear fuel (SNF). There are two basic strategies of SNF management: ultimate disposal of SNF in geological formations and recycle or repeated utilisation of reprocessed SNF. In both options, there is an urgent necessity to study radiation properties of SNF. Information about SNF radiation properties is required at all stages of SNF management. In order to reach more effective utilisation of nuclear materials, new fuel cycles are under development based on uranium-plutonium, uranium-thorium and some other types of nuclear fuel. These promising types of nuclear fuel are characterised by quite different radiation properties at all the stages of nuclear fuel cycle (NFC) listed above. So, comparative analysis is required for radiation properties of different nuclear fuel types at different NFC stages. The results presented here were obtained from the numerical analysis of the radiation field around transport containers of different SNF types and in SNF storage. The calculations are carried out with the application of the computer code packages SCALE-4.3 and MCNP-4C. Comparison of the dose parameters obtained for different models of the transport container with experimental data allowed us to make certain conclusions about the errors of numerical results caused by the approximate geometrical description of the transport container.

MTR BUILDING AND BALCONY FLOORS. CAMERA FACING EASTERLY. PHOTOGRAPHER DID ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR BUILDING AND BALCONY FLOORS. CAMERA FACING EASTERLY. PHOTOGRAPHER DID NOT EXPLAIN DARK CLOUD. MTR WING WILL ATTACH TO GROUND FLOOR. INL NEGATIVE NO. 1567. Unknown Photographer, 2/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Characterization of the radiation environment for a large-area interim spent-nuclear-fuel storage facility

NASA Astrophysics Data System (ADS)

Fortkamp, Jonathan C.

Current needs in the nuclear industry and movements in the political arena indicate that authorization may soon be given for development of a federal interim storage facility for spent nuclear fuel. The initial stages of the design work have already begun within the Department of Energy and are being reviewed by the Nuclear Regulatory Commission. This dissertation addresses the radiation environment around an interim spent nuclear fuel storage facility. Specifically the dissertation characterizes the radiation dose rates around the facility based on a design basis source term, evaluates the changes in dose due to varying cask spacing configurations, and uses these results to define some applicable health physics principles for the storage facility. Results indicate that dose rates from the facility are due primarily from photons from the spent fuel and Co-60 activation in the fuel assemblies. In the modeled cask system, skyshine was a significant contribution to dose rates at distances from the cask array, but this contribution can be reduced with an alternate cask venting system. With the application of appropriate health physics principles, occupation doses can be easily maintained far below regulatory limits and maintained ALARA.

CONTEXTUAL AERIAL VIEW OF "EXCLUSION" MTR AREA WITH IDAHO CHEMICAL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

CONTEXTUAL AERIAL VIEW OF "EXCLUSION" MTR AREA WITH IDAHO CHEMICAL PROCESSING PLANT IN BACKGROUND AT CENTER TOP OF VIEW. CAMERA FACING EAST. EXCLUSION GATE HOUSE AT LEFT OF VIEW. BEYOND MTR BUILDING AND ITS WING, THE PROCESS WATER BUILDING AND WORKING RESERVOIR ARE LEFT-MOST. FAN HOUSE AND STACK ARE TO ITS RIGHT. PLUG STORAGE BUILDING IS RIGHT-MOST STRUCTURE. NOTE FAN LOFT ABOVE MTR BUILDING'S ONE-STORY WING. THIS WAS LATER CONVERTED FOR OFFICES. INL NEGATIVE NO. 3610. Unknown Photographer, 10/30/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

A Specific Long-Term Plan for Management of U.S. Nuclear Spent Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Levy, Salomon

2006-07-01

A specific plan consisting of six different steps is proposed to accelerate and improve the long-term management of U.S. Light Water Reactor (LWR) spent nuclear fuel. The first step is to construct additional, centralized, engineered (dry cask) spent fuel facilities to have a backup solution to Yucca Mountain (YM) delays or lack of capacity. The second step is to restart the development of the Integral Fast Reactor (IFR), in a burner mode, because of its inherent safety characteristics and its extensive past development in contrast to Acceleration Driven Systems (ADS). The IFR and an improved non-proliferation version of its pyro-processingmore » technology can burn the plutonium (Pu) and minor actinides (MA) obtained by reprocessing LWR spent fuel. The remaining IFR and LWR fission products will be treated for storage at YM. The radiotoxicity of that high level waste (HLW) will fall below that of natural uranium in less than one thousand years. Due to anticipated increased capital, maintenance, and research costs for IFR, the third step is to reduce the required number of IFRs and their potential delays by implementing multiple recycles of Pu and Neptunium (Np) MA in LWR. That strategy is to use an advanced separation process, UREX+, and the MIX Pu option where the role and degradation of Pu is limited by uranium enrichment. UREX+ will decrease proliferation risks by avoiding Pu separation while the MIX fuel will lead to an equilibrium fuel recycle mode in LWR which will reduce U. S. Pu inventory and deliver much smaller volumes of less radioactive HLW to YM. In both steps two and three, Research and Development (R and D) is to emphasize the demonstration of multiple fuel reprocessing and fabrication, while improving HLW treatment, increasing proliferation resistance, and reducing losses of fissile material. The fourth step is to license and construct YM because it is needed for the disposal of defense wastes and the HLW to be generated under the proposed plan

Spent nuclear fuel canister storage building conceptual design report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swenson, C.E.

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

10 CFR 72.214 - List of approved spent fuel storage casks.

Code of Federal Regulations, 2012 CFR

2012-01-01

... specified in their Certificates of Compliance. Certificate Number: 1000 SAR Submitted by: General Nuclear Systems, Inc. SAR Title: Topical Safety Analysis Report for the Castor V/21 Cask Independent Spent Fuel... Model Number: CASTOR V/21 Certificate Number: 1002 SAR Submitted by: Nuclear Assurance Corporation SAR...

10 CFR 72.214 - List of approved spent fuel storage casks.

Code of Federal Regulations, 2014 CFR

2014-01-01

... specified in their Certificates of Compliance. Certificate Number: 1000 SAR Submitted by: General Nuclear Systems, Inc. SAR Title: Topical Safety Analysis Report for the Castor V/21 Cask Independent Spent Fuel... Model Number: CASTOR V/21 Certificate Number: 1002 SAR Submitted by: Nuclear Assurance Corporation SAR...

10 CFR 72.214 - List of approved spent fuel storage casks.

Code of Federal Regulations, 2013 CFR

2013-01-01

... specified in their Certificates of Compliance. Certificate Number: 1000 SAR Submitted by: General Nuclear Systems, Inc. SAR Title: Topical Safety Analysis Report for the Castor V/21 Cask Independent Spent Fuel... Model Number: CASTOR V/21 Certificate Number: 1002 SAR Submitted by: Nuclear Assurance Corporation SAR...

Monte Carlo characterization of PWR spent fuel assemblies to determine the detectability of pin diversion

NASA Astrophysics Data System (ADS)

Burdo, James S.

This research is based on the concept that the diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies is feasible by a careful comparison of spontaneous fission neutron and gamma levels in the guide tube locations of the fuel assemblies. The goal is to be able to determine whether some of the assembly fuel pins are either missing or have been replaced with dummy or fresh fuel pins. It is known that for typical commercial power spent fuel assemblies, the dominant spontaneous neutron emissions come from Cm-242 and Cm-244. Because of the shorter half-life of Cm-242 (0.45 yr) relative to that of Cm-244 (18.1 yr), Cm-244 is practically the only neutron source contributing to the neutron source term after the spent fuel assemblies are more than two years old. Initially, this research focused upon developing MCNP5 models of PWR fuel assemblies, modeling their depletion using the MONTEBURNS code, and by carrying out a preliminary depletion of a ¼ model 17x17 assembly from the TAKAHAMA-3 PWR. Later, the depletion and more accurate isotopic distribution in the pins at discharge was modeled using the TRITON depletion module of the SCALE computer code. Benchmarking comparisons were performed with the MONTEBURNS and TRITON results. Subsequently, the neutron flux in each of the guide tubes of the TAKAHAMA-3 PWR assembly at two years after discharge as calculated by the MCNP5 computer code was determined for various scenarios. Cases were considered for all spent fuel pins present and for replacement of a single pin at a position near the center of the assembly (10,9) and at the corner (17,1). Some scenarios were duplicated with a gamma flux calculation for high energies associated with Cm-244. For each case, the difference between the flux (neutron or gamma) for all spent fuel pins and with a pin removed or replaced is calculated for each guide tube. Different detection criteria were established. The first was whether the relative error of the

78 FR 73456 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-06

...-2012-0052] RIN 3150-AJ12 List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment... International HI-STORM 100 Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to... requirements for the HI-STORM 100U part of the HI-STORM 100 Cask System and updates the thermal model and...

Radiolytic and Thermal Process Relevant to Dry Storage of Spent Nuclear Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marschman, Steven C.; Haustein, Peter E.; Madey, Theodore E.

1999-06-01

This project involves basic research in chemistry and physics aimed at providing information pertinent to the safe long-term dry storage of spent nuclear fuel (SNF), thousands of tons of which remain in water storage across the DOE complex. The Hanford Site K-Basins alone hold 2300 tons of spent fuel, much of it severely corroded, and similar situations exist at Savannah River and Idaho National Engineering and Environmental Laboratory. DOE plans to remove this fuel and seal it in overpack canisters for ''dry'' interim storage for up to 75 years while awaiting permanent disposition. Chemically bound water will remain in thismore » fuel even after the proposed drying steps, leading to possible long-term corrosion of the containers and/or fuel rods themselves, generation of H2 and O2 gas via radiolysis (which could lead to deflagration or detonation), and reactions of pyrophoric uranium hydrides. No thoroughly tested model is now available to predict fuel behavior during preprocessing, processing, or storage. In a collaborative effort among Rutgers University, Pacific Northwest National Laboratory, and Brookhaven National Laboratory, we are studying the radiolytic reaction, drying processes, and corrosion behavior of actual SNF materials and of pure and mixed-phase samples. We propose to determine what is omitted from current models: radiolysis of water adsorbed on or in hydrates or hydroxides, thermodynamics of interfacial phases, and kinetics of drying. A model will be developed and tested against actual fuel rod behavior to ensure validity and applicability to the problems associated with developing dry storage strategies for DOE-owned SNF.« less

REACTIVITY MEASUREMENT FACILITY, UNDER CONSTRUCTION OVER MTR CANAL IN BASEMENT ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

REACTIVITY MEASUREMENT FACILITY, UNDER CONSTRUCTION OVER MTR CANAL IN BASEMENT OF MTR BUILDING, TRA-603. WOOD PLANKS REST ON CANAL WALL OBSERVABLE IN FOREGROUND. INL NEGATIVE NO. 11745. Unknown Photographer, 8/20/1954 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Mechanical and thermomechanical calculations related to the storage of spent nuclear-fuel assemblies in granite

NASA Astrophysics Data System (ADS)

Butkovich, T. R.

1981-08-01

A generic test of the geologic storage of spent-fuel assemblies from an operating nuclear reactor is being made by the Lawrence Livermore National Laboratory at the US Department of Energy's Nevada Test Site. The spent-fuel assemblies were emplaced at a depth of 420 m (1370 ft) below the surface in a typical granite and will be retrieved at a later time. The early time, close-in thermal history of this type of repository is being simulated with spent-fuel and electrically heated canisters in a central drift, with auxiliary heaters in two parallel side drifts. Prior to emplacement of the spent-fuel canister, preliminary calculations were made using a pair of existing finite-element codes. Calculational modeling of a spent-fuel repository requires a code with a multiple capability. The effects of both the mining operation and the thermal load on the existing stress fields and the resultant displacements of the rock around the repository must be calculated. The thermal loading for each point in the rock is affected by heat transfer through conduction, radiation, and normal convection, as well as by ventilation of the drifts. Both the ADINA stress code and the compatible ADINAT heat-flow code were used to perform the calculations because they satisfied the requirements of this project. ADINAT was adapted to calculate radiative and convective heat transfer across the drifts and to model the effects of ventilation in the drifts, while the existing isotropic elastic model was used with the ADINA code. The results of the calculation are intended to provide a base with which to compare temperature, stress, and displacement data taken during the planned 5-y duration of the test. In this way, it will be possible to determine how the existing jointing in the rock influences the results as compared with a homogeneous, isotropic rock mass. Later, new models will be introduced into ADINA to account for the effects of jointing.

10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

Code of Federal Regulations, 2010 CFR

2010-01-01

..., reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design Criteria § 72.128 Criteria for spent fuel, high-level radioactive waste, reactor...

10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

Code of Federal Regulations, 2011 CFR

2011-01-01

..., reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design Criteria § 72.128 Criteria for spent fuel, high-level radioactive waste, reactor...

Radiolytic and Thermal Processes Relevant to Dry Storage of Spent Nuclear Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marschman, Steven C.; Madey,Theodore E.; Haustein, Peter E.

2000-06-01

The purpose of this project is to deliver pertinent information that can be used to make rational decisions about the safety and treatment issues associated with dry storage of spent nuclear fuel materials. In particular, we will establish an understanding of: (1) water interactions with failed-fuel rods and metal-oxide materials; (2) the role of thermal processes and radiolysis (solid-state and interfacial) in the generation of potentially explosive mixtures of gaseous H2 and O2; and (3) the potential role of radiation-assisted corrosion during fuel rod storage.

Method for reprocessing and separating spent nuclear fuels. [Patent application

DOEpatents

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

1982-01-19

Spent nuclear fuels, including actinide fuels, volatile and nonvolatile fission products, are reprocessed and separated in a molten metal solvent housed in a separation vessel made of a carbon-containing material. A first catalyst, which promotes the solubility and permeability of carbon in the metal solvent, is included. By increasing the solubility and permeability of the carbon in the solvent, the rate at which actinide oxides are reduced (carbothermic reduction) is greatly increased. A second catalyst, included to increase the affinity for nitrogen in the metal solvent, is added to increase the rate at which actinide nitrides form after carbothermic reduction is complete.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Newsom, H.C.

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

76 FR 34007 - Draft Regulatory Basis for a Potential Rulemaking on Spent Nuclear Fuel Reprocessing Facilities

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-10

... processes are more akin to fuel cycle processes. This framework was established in the 1970's to license the... nuclear power globally and close the nuclear fuel cycle through reprocessing spent fuel and deploying fast... Accounting;'' and a Nuclear Energy Institute white [[Page 34009

Corrosion property of 9Cr-ODS steel in nitric acid solution for spent nuclear fuel reprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takeuchi, M.; Koizumi, T.; Inoue, M.

2013-07-01

Corrosion tests of oxide dispersion strengthened with 9% Cr (9Cr-ODS) steel, which is one of the desirable materials for cladding tube of sodium-cooled fast reactors, in pure nitric acid solution, spent FBR fuel solution, and its simulated solution were performed to understand the corrosion behavior in a spent nuclear fuel reprocessing. In this study, the 9Cr-ODS steel with lower effective chromium content was evaluated to understand the corrosion behavior conservatively. As results, the tube-type specimens of the 9Cr-ODS steels suffered severe weight loss owing to active dissolution at the beginning of the immersion test in pure nitric acid solution inmore » the range from 1 to 3.5 M. In contrast, the weight loss was decreased and they showed a stable corrosion in the higher nitric acid concentration, the dissolved FBR fuel solution, and its simulated solution by passivation. The corrosion rates of the 9Cr-ODS steel in the dissolved FBR fuel solution and its simulated solution were 1-2 mm/y and showed good agreement with each other. The passivation was caused by the shift of corrosion potential to noble side owing to increase in nitric acid concentration or oxidative ions in the dissolved FBR fuel solution and the simulated spent fuel solution. (authors)« less

Assessing the effectiveness of safeguards at a medium-sized spent-fuel reprocessing facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Higinbotham, W.; Fishbone, L.G.; Suda, S.

1983-01-01

In order to evaluate carefully and systematically the effectiveness of safeguards at nuclear-fuel-cycle facilities, the International Atomic Energy Agency has adopted a safeguards effectiveness assessment methodology. The methodology has been applied to a well-characterized, medium-sized, spent-fuel reprocessing plant to understand how explicit safeguards inspection procedures would serve to expose conceivable nuclear materials diversion schemes, should such diversion occur.

Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

DOEpatents

Ackerman, John P.; Miller, William E.

1989-01-01

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuel using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuel, and two cathodes, the first cathode composed of either a solid alloy or molten cadmium and the second cathode composed of molten cadmium. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then substantially pure uranium is electrolytically transported and deposited on the first alloy or molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on the second molten cadmium cathode.

Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

DOEpatents

Ackerman, J.P.; Miller, W.E.

1987-11-05

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuels is disclosed using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuels, two cathodes and electrical power means connected to the anode basket, cathodes and lower molten cadmium pool for providing electrical power to the cell. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then purified uranium is electrolytically transported and deposited on a first molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on a second cathode. 3 figs.

MELCOR model for an experimental 17x17 spent fuel PWR assembly.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cardoni, Jeffrey

2010-11-01

A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

75 FR 9452 - Solicitation of Topics for Discussion at a Spent Fuel Storage and Transportation Licensing...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-02

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0076] Solicitation of Topics for Discussion at a Spent...: Solicitation of Topics for Discussion at a Spent Fuel Storage and Transportation Licensing Conference. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is soliciting input on topics for discussion at a...

78 FR 16601 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-18

... Storage Casks: MAGNASTOR[supreg] System AGENCY: Nuclear Regulatory Commission. ACTION: Direct final rule. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is amending its spent fuel storage regulations by revising the NAC International, Inc. (NAC) Modular Advanced Generation Nuclear All-purpose Storage...

PRECONSTRUCTION IMAGE OF THE MTR SITE. ABANDONED IRRIGATION CANAL (FROM ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PRE-CONSTRUCTION IMAGE OF THE MTR SITE. ABANDONED IRRIGATION CANAL (FROM EARLY 1900s) ILLUSTRATES FLATNESS OF MTR/TRA TERRAIN. FEATURE ON HORIZON IN LEFT OF VIEW IS EXPLORATORY WATER DRILLING EQUIPMENT. CAMERA LOOKS SOUTHEAST. INL NEGATIVE NO. 136. Unknown Photographer, 12/5/1949 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koester, L.W.

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

75 FR 27463 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1; Correction

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-17

... Fuel Storage Casks: NUHOMS[supreg] HD System Revision 1; Correction AGENCY: Nuclear Regulatory... fuel storage casks to add revision 1 to the NUHOMS HD spent fuel storage cask system. This action is... Federal Register on May 7, 2010 (75 FR 25120), that proposes to amend the regulations that govern storage...

Spent nuclear fuel system dynamic stability under normal conditions of transportation

DOE PAGES

Jiang, Hao; Wang, Jy-An John

2016-10-14

In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside themore » cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.« less

Spent nuclear fuel system dynamic stability under normal conditions of transportation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, Hao; Wang, Jy-An John

In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside themore » cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.« less

Spent fuel and high-level radioactive waste transportation report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educatedmore » layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages sew be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.« less

Spent fuel and high-level radioactive waste transportation report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educatedmore » layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.« less

Spent Fuel and High-Level Radioactive Waste Transportation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by SSEB in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or nomore » background in nuclear waste Issues. In addition. this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.« less

Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

DOE PAGES

Favalli, Andrea; Vo, D.; Grogan, Brandon R.; ...

2016-02-26

The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/ 137Cs, 134Cs/ 137Cs, 106Ru/ 137Cs, and 144Ce/ 137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

DOE Office of Scientific and Technical Information (OSTI.GOV)

Favalli, Andrea; Vo, D.; Grogan, Brandon R.

The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/ 137Cs, 134Cs/ 137Cs, 106Ru/ 137Cs, and 144Ce/ 137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

Spent fuel treatment and mineral waste form development at Argonne National Laboratory-West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goff, K.M.; Benedict, R.W.; Bateman, K.

1996-07-01

At Argonne National Laboratory-West (ANL-West) there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Conditioning Facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. Both mineral and metal high-level waste forms will be produced. The mineral waste form will contain the active metal fission products and the transuranics. Cold small-scale waste form testing has been on-going at Argonne in Illinois. Large-scale testing is commencing at ANL-West.

Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

NASA Astrophysics Data System (ADS)

Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

2017-09-01

Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sheryl L. Morton; Philip L. Winston; Toshiari Saegusa

2006-04-01

In 2003, representatives from the Central Research Institute of Electric Power Industry (CRIEPI) requested development of a project with the objective of determining the performance of a concrete spent nuclear fuel storage cask. Radiation and environmental effects may cause chemical alteration of the concrete that could result in excessive cracking, spalling, and loss of compressive strength. The Idaho National Laboratory (INL) project team and CRIEPI representatives identified the Ventilated Storage Cask (VSC-17) spent nuclear fuel storage cask as a candidate to study cask performance, because it had been used to store fuel as part of a dry cask storage demonstrationmore » project for more than 15 years. The project involved investigating the properties of the concrete shield. INL performed a survey of the cask in the summers of 2003 and 2004. Preliminary cask evaluations performed in 2003 indicated that the cask has no visual degradation. However, a 4-5 mrem/hr step-change in the radiation levels about halfway up the cask and a localized hot spot beneath an upper air vent indicate that there may be variability in the density of the concrete or localized cracking. In 2005, INL and CRIEPI scientists performed additional surveys on the VSC-17 cask. This document summarizes the methods used on the VSC-17 to evaluate the cask for compressive strength, concrete cracking, concrete thickness, and temperature distribution.« less

Update and evaluation of decay data for spent nuclear fuel analyses

NASA Astrophysics Data System (ADS)

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.

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

NASA Astrophysics Data System (ADS)

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

2013-09-01

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

Creating NDA working standards through high-fidelity spent fuel modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skutnik, Steven E; Gauld, Ian C; Romano, Catherine E

2012-01-01

The Next Generation Safeguards Initiative (NGSI) is developing advanced non-destructive assay (NDA) techniques for spent nuclear fuel assemblies to advance the state-of-the-art in safeguards measurements. These measurements aim beyond the capabilities of existing methods to include the evaluation of plutonium and fissile material inventory, independent of operator declarations. Testing and evaluation of advanced NDA performance will require reference assemblies with well-characterized compositions to serve as working standards against which the NDA methods can be benchmarked and for uncertainty quantification. To support the development of standards for the NGSI spent fuel NDA project, high-fidelity modeling of irradiated fuel assemblies is beingmore » performed to characterize fuel compositions and radiation emission data. The assembly depletion simulations apply detailed operating history information and core simulation data as it is available to perform high fidelity axial and pin-by-pin fuel characterization for more than 1600 nuclides. The resulting pin-by-pin isotopic inventories are used to optimize the NDA measurements and provide information necessary to unfold and interpret the measurement data, e.g., passive gamma emitters, neutron emitters, neutron absorbers, and fissile content. A key requirement of this study is the analysis of uncertainties associated with the calculated compositions and signatures for the standard assemblies; uncertainties introduced by the calculation methods, nuclear data, and operating information. An integral part of this assessment involves the application of experimental data from destructive radiochemical assay to assess the uncertainty and bias in computed inventories, the impact of parameters such as assembly burnup gradients and burnable poisons, and the influence of neighboring assemblies on periphery rods. This paper will present the results of high fidelity assembly depletion modeling and uncertainty analysis from

78 FR 39781 - Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-02

... Spent Fuel Pool Study). The purpose of this study was to examine if faster removal of older, colder... NRC Library at http://www.nrc.gov/reading-rm/adams.html . To begin the search, select ``ADAMS Public... of postulated spent fuel pool accidents. The purpose of this study is to examine if faster removal of...

78 FR 73566 - Standard Format and Content for a License Application for an Independent Spent Fuel Storage...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-06

...The U.S. Nuclear Regulatory Commission (NRC) is issuing for public comment draft regulatory guide (DG), DG-3042, ``Standard Format and Content for a License Application for an Independent Spent Fuel Storage Installation or a Monitored Retrievable Storage Facility.'' This draft regulatory guide is proposed revision 2 of Regulatory Guide 3.50, which provides a format that the NRC considers acceptable for submitting the information for license applications to store spent nuclear fuel, high-level radioactive waste, and/or reactor-related Greater than Class C waste.

Reforming of fuel inside fuel cell generator

DOEpatents

Grimble, Ralph E.

1988-01-01

Disclosed is an improved method of reforming a gaseous reformable fuel within a solid oxide fuel cell generator, wherein the solid oxide fuel cell generator has a plurality of individual fuel cells in a refractory container, the fuel cells generating a partially spent fuel stream and a partially spent oxidant stream. The partially spent fuel stream is divided into two streams, spent fuel stream I and spent fuel stream II. Spent fuel stream I is burned with the partially spent oxidant stream inside the refractory container to produce an exhaust stream. The exhaust stream is divided into two streams, exhaust stream I and exhaust stream II, and exhaust stream I is vented. Exhaust stream II is mixed with spent fuel stream II to form a recycle stream. The recycle stream is mixed with the gaseous reformable fuel within the refractory container to form a fuel stream which is supplied to the fuel cells. Also disclosed is an improved apparatus which permits the reforming of a reformable gaseous fuel within such a solid oxide fuel cell generator. The apparatus comprises a mixing chamber within the refractory container, means for diverting a portion of the partially spent fuel stream to the mixing chamber, means for diverting a portion of exhaust gas to the mixing chamber where it is mixed with the portion of the partially spent fuel stream to form a recycle stream, means for injecting the reformable gaseous fuel into the recycle stream, and means for circulating the recycle stream back to the fuel cells.

Reforming of fuel inside fuel cell generator

DOEpatents

Grimble, R.E.

1988-03-08

Disclosed is an improved method of reforming a gaseous reformable fuel within a solid oxide fuel cell generator, wherein the solid oxide fuel cell generator has a plurality of individual fuel cells in a refractory container, the fuel cells generating a partially spent fuel stream and a partially spent oxidant stream. The partially spent fuel stream is divided into two streams, spent fuel stream 1 and spent fuel stream 2. Spent fuel stream 1 is burned with the partially spent oxidant stream inside the refractory container to produce an exhaust stream. The exhaust stream is divided into two streams, exhaust stream 1 and exhaust stream 2, and exhaust stream 1 is vented. Exhaust stream 2 is mixed with spent fuel stream 2 to form a recycle stream. The recycle stream is mixed with the gaseous reformable fuel within the refractory container to form a fuel stream which is supplied to the fuel cells. Also disclosed is an improved apparatus which permits the reforming of a reformable gaseous fuel within such a solid oxide fuel cell generator. The apparatus comprises a mixing chamber within the refractory container, means for diverting a portion of the partially spent fuel stream to the mixing chamber, means for diverting a portion of exhaust gas to the mixing chamber where it is mixed with the portion of the partially spent fuel stream to form a recycle stream, means for injecting the reformable gaseous fuel into the recycle stream, and means for circulating the recycle stream back to the fuel cells. 1 fig.

77 FR 9591 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-17

... Fuel Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear Regulatory Commission. ACTION: Proposed... spent fuel storage cask regulations by revising the Holtec International HI-STORM 100 dry cask storage... Amendment No. 8 to CoC No. 1014 and does not include other aspects of the HI-STORM 100 dry storage cask...

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

NASA Astrophysics Data System (ADS)

Durham, J. M.; Poulson, D.; Bacon, J.; Chichester, D. L.; Guardincerri, E.; Morris, C. L.; Plaud-Ramos, K.; Schwendiman, W.; Tolman, J. D.; Winston, P.

2018-04-01

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. Here we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. This application of technology and methods commonly used in high-energy particle physics provides a potential solution to this long-standing problem in international nuclear safeguards.

76 FR 35137 - Vulnerability and Threat Information for Facilities Storing Spent Nuclear Fuel and High-Level...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-16

... High-Level Radioactive Waste AGENCY: U.S. Nuclear Regulatory Commission. ACTION: Public meeting... Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste,'' and 73... Spent Nuclear Fuel (SNF) and High-Level Radioactive Waste (HLW) storage facilities. The draft regulatory...

MTR MAIN FLOOR. MEN DEMONSTRATE INSERTION OF DUMMY PLUG INTO ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR MAIN FLOOR. MEN DEMONSTRATE INSERTION OF DUMMY PLUG INTO AN MTR BEAM HOLE. ONE MAN CHECKS RADIATION LEVEL AT THE END OF THE UNIVERSAL COFFIN, WHILE ANOTHER USES TOOL TO INSERT PLUG INTO HOLE THROUGH COFFIN. MEN WEAR "ANTI-C" (ANTI-CONTAMINATION) CLOTHING. INL NEGATIVE NO. 6198. R.G. Larsen, Photographer, 6/27/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Strategic Minimization of High Level Waste from Pyroprocessing of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simpson, Michael F.; Benedict, Robert W.

The pyroprocessing of spent nuclear fuel results in two high-level waste streams--ceramic and metal waste. Ceramic waste contains active metal fission product-loaded salt from the electrorefining, while the metal waste contains cladding hulls and undissolved noble metals. While pyroprocessing was successfully demonstrated for treatment of spent fuel from Experimental Breeder Reactor-II in 1999, it was done so without a specific objective to minimize high-level waste generation. The ceramic waste process uses “throw-away” technology that is not optimized with respect to volume of waste generated. In looking past treatment of EBR-II fuel, it is critical to minimize waste generation for technologymore » developed under the Global Nuclear Energy Partnership (GNEP). While the metal waste cannot be readily reduced, there are viable routes towards minimizing the ceramic waste. Fission products that generate high amounts of heat, such as Cs and Sr, can be separated from other active metal fission products and placed into short-term, shallow disposal. The remaining active metal fission products can be concentrated into the ceramic waste form using an ion exchange process. It has been estimated that ion exchange can reduce ceramic high-level waste quantities by as much as a factor of 3 relative to throw-away technology.« less

Analyses of Field Test Data at the Atucha-1 Spent Fuel Pools

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sitaraman, S.

A field test was conducted at the Atucha-1 spent nuclear fuel pools to validate a software package for gross defect detection that is used in conjunction with the inspection tool, Spent Fuel Neutron Counter (SFNC). A set of measurements was taken with the SFNC and the software predictions were compared with these data and analyzed. The data spanned a wide range of cooling times and a set of burnup levels leading to count rates from the several hundreds to around twenty per second. The current calibration in the software using linear fitting required the use of multiple calibration factors tomore » cover the entire range of count rates recorded. The solution to this was to use power regression data fitting to normalize the predicted response and derive one calibration factor that can be applied to the entire set of data. The resulting comparisons between the predicted and measured responses were generally good and provided a quantitative method of detecting missing fuel in virtually all situations. Since the current version of the software uses the linear calibration method, it would need to be updated with the new power regression method to make it more user-friendly for real time verification and fieldable for the range of responses that will be encountered.« less

Determination of total plutonium content in spent nuclear fuel assemblies with the differential die-away self-interrogation instrument

NASA Astrophysics Data System (ADS)

Kaplan, Alexis C.; Henzl, Vladimir; Menlove, Howard O.; Swinhoe, Martyn T.; Belian, Anthony P.; Flaska, Marek; Pozzi, Sara A.

2014-11-01

As a part of the Next Generation Safeguards Initiative Spent Fuel project, we simulate the response of the Differential Die-away Self-Interrogation (DDSI) instrument to determine total elemental plutonium content in an assayed spent nuclear fuel assembly (SFA). We apply recently developed concepts that relate total plutonium mass with SFA multiplication and passive neutron count rate. In this work, the multiplication of the SFA is determined from the die-away time in the early time domain of the Rossi-Alpha distributions measured directly by the DDSI instrument. We utilize MCNP to test the method against 44 pressurized water reactor SFAs from a simulated spent fuel library with a wide dynamic range of characteristic parameters such as initial enrichment, burnup, and cooling time. Under ideal conditions, discounting possible errors of a real world measurement, a root mean square agreement between true and determined total Pu mass of 2.1% is achieved.

Mechanical tissue resuscitation (MTR): a nonpharmacological approach to treatment of acute myocardial infarction.

PubMed

Jordan, James E; Pereira, Beatriz D; Lane, Magan R; Morykwas, Michael J; McGee, Maria; Argenta, Louis C

2015-08-01

Myocardial ischemia-reperfusion injury is known to trigger an inflammatory response involving edema, apoptosis, and neutrophil activation/accumulation. Recently, mechanical tissue resuscitation (MTR) was described as a potent cardioprotective strategy for reduction of myocardial ischemia-reperfusion injury. Here, we further describe the protective actions of MTR and begin to define its therapeutic window. A left ventricular, free-wall ischemic area was created in anesthetized swine for 85 minutes and then reperfused for three hours. Animals were randomized to two groups: (1) untreated controls (Control) and (2) application of MTR that was delayed 90 minutes after the initiation of reperfusion (D90). Hemodynamics and regional myocardial blood flow were assessed at multiple time points. Infarct size and neutrophil accumulation were assessed following the reperfusion period. In separate cohorts, the effect of MTR on myocardial interstitial water (MRI imaging) and blood flow was examined. Both groups had similar areas at risk (AAR), hemodynamics, and arterial blood gas values. MTR, even when delayed 90 minutes into reperfusion (D90, 29.2 ± 5.0% of AAR), reduced infarct size significantly compared to Controls (51.9 ± 2.7%, p = 0.006). This protection was associated with a 33% decrease in neutrophil accumulation (p = 0.047). Improvements in blood flow and interstitial water were also observed. Moreover, we demonstrated that the therapeutic window for MTR lasts for at least 90 minutes following reperfusion. This study confirms our previous observations that MTR is an effective therapeutic approach to reducing reperfusion injury with a clinically useful treatment window. © 2015 Wiley Periodicals, Inc.

Spent nuclear fuel recycling with plasma reduction and etching

DOEpatents

Kim, Yong Ho

2012-06-05

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

NRC approves spent-fuel cask for general use: Who needs Yucca Mountain?

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simpson, J.

1993-07-01

The Nuclear Regulatory Commission (NRC) on April 7, 1993, added Pacific Sierra Nuclear Associates`s (PSNA`s) VSC-24 spent-fuel container to its list of approved storage casks. Unlike previously approved designs, however, the cask was made available for use by utilities without site-specific approval. The VSC-24 (ventilated storage cask) is a 130-ton, 16-foot high vertical storage container composed of a ventilated concrete cask (VCC) housing a steel multi-assembly sealed basket (MSB). A third component, a transfer cask (MTC), shields, supports, and protects the MSB during fuel loading and VCC loading operations. The VCC is a cylindrical reinforced-concrete cask 29 inches thick, withmore » a 1.75-inch-thick A 36 steel liner. The cask contains eight vents-four on the top and four on the bottom-to provide for MSB (and fuel rod) cooling. Its concrete shell provides protection against shearing and penetration by tornado projectiles, protects the MSB in the event of a drop or tipover, and is designed to withstand internal temperatures of 350 degrees Farenheit. The VCC is closed with a bolted-down cover of 0.75-inch-thick A 36 steel. The MSB, which provides the primary boundary for 24 spent fuel rods, is a cylindrical steel shell with a thick shield plug and steel cover plates welded at each end. The shell and covers are constructed from SA 516 Grade 70 pressure vessel steel. Fuel is housed in a basket fabricated from SA 516 Grade 70 sheet steel. Penetrations in the MSB`s structural and shield lids allow for vacuum drying and backfilling with helium after fuel loading. Although its manufacturer claims a design life of 50 years, the NRC has licensed the VSC-24 cask for 20 years.« less

Production of a solid fuel using sewage sludge and spent cooking oil by immersion frying.

PubMed

Wu, Zhonghua; Zhang, Jing; Li, Zhanyong; Xie, Jian; Mujumdar, Arun S

2012-12-01

Sewage sludge and spent cooking oil are two main waste sources of modern Chinese cities. In this paper, the immersion frying method using spent cooking oil as the heating medium was applied to dry and convert wet sewage sludge into a solid fuel. The drying and oil uptake curves were plotted to demonstrate the fry-drying characteristics of the sewage sludge. Parametric studies were carried out to identify the governing parameters in the frying drying operation. It was found that at frying oil temperatures of 140-160°C, the wet sewage sludge could be dried completely in 6-9 min and converted into a solid fuel with a high calorific value of 21.55-24.08 MJ/kg. The fuel structure, chemical components, pyrolysis and combustion characteristics were investigated and the experimental results showed the solid fuel had a porous internal structure and a low ignition temperature of 250°C due to presence of oil. The frying drying mechanism was also discussed. Copyright © 2012 Elsevier B.V. All rights reserved.

Global Spent Fuel Logistics Systems Study (GSFLS). Volume 2A. GSFLS visit findings (appendix). Interim report

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1978-01-31

This appendix is a part of the interim report documentation for the Global Spent Fuel Logistics System (GSFLS) study. This appendix provides the legal/regulatory reference material, supportive of Volume 2 - GSFLS Visit Finding and Evaluations; and certain background material on British Nuclear Fuel Limited (BNFL).

Nuclear Nonproliferation: Concerns With U.S. Delays in Accepting Foreign Research Reactors’ Spent Fuel

DTIC Science & Technology

1994-03-01

transport or storage plans. The return of some of the spent fuel will also depend on the readiness of dry storage . One expert told us that...enriched uranium fuel (HEU), a material that can be used to make nuclear bombs, in civilian nuclear programs worldwide. Research reactors are of...address the environmental impact of transporting the fuel and storing it in both existing and new storage units, possibly by June 1995. Under the

Cosmic Ray Muon Imaging of Spent Nuclear Fuel in Dry Storage Casks

DOE PAGES

Durham, J. Matthew; Guardincerri, Elena; Morris, Christopher L.; ...

2016-04-29

In this paper, cosmic ray muon radiography has been used to identify the absence of spent nuclear fuel bundles inside a sealed dry storage cask. The large amounts of shielding that dry storage casks use to contain radiation from the highly radioactive contents impedes typical imaging methods, but the penetrating nature of cosmic ray muons allows them to be used as an effective radiographic probe. This technique was able to successfully identify missing fuel bundles inside a sealed Westinghouse MC-10 cask. This method of fuel cask verification may prove useful for international nuclear safeguards inspectors. Finally, muon radiography may findmore » other safety and security or safeguards applications, such as arms control verification.« less

Thermoelastic analysis of spent fuel and high level radioactive waste repositories in salt. A semi-analytical solution. [JUDITH

DOE Office of Scientific and Technical Information (OSTI.GOV)

St. John, C.M.

1977-04-01

An underground repository containing heat generating, High Level Waste or Spent Unreprocessed Fuel may be approximated as a finite number of heat sources distributed across the plane of the repository. The resulting temperature, displacement and stress changes may be calculated using analytical solutions, providing linear thermoelasticity is assumed. This report documents a computer program based on this approach and gives results that form the basis for a comparison between the effects of disposing of High Level Waste and Spent Unreprocessed Fuel.

Container for reprocessing and permanent storage of spent nuclear fuel assemblies

DOEpatents

Forsberg, Charles W.

1992-01-01

A single canister process container for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining their integrity at temperature necessary to oxide the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container.

MtrA of the sodium ion pumping methyltransferase binds cobalamin in a unique mode

PubMed Central

Wagner, Tristan; Ermler, Ulrich; Shima, Seigo

2016-01-01

In the three domains of life, vitamin B12 (cobalamin) is primarily used in methyltransferase and isomerase reactions. The methyltransferase complex MtrA–H of methanogenic archaea has a key function in energy conservation by catalysing the methyl transfer from methyl-tetrahydromethanopterin to coenzyme M and its coupling with sodium-ion translocation. The cobalamin-binding subunit MtrA is not homologous to any known B12-binding proteins and is proposed as the motor of the sodium-ion pump. Here, we present crystal structures of the soluble domain of the membrane-associated MtrA from Methanocaldococcus jannaschii and the cytoplasmic MtrA homologue/cobalamin complex from Methanothermus fervidus. The MtrA fold corresponds to the Rossmann-type α/β fold, which is also found in many cobalamin-containing proteins. Surprisingly, the cobalamin-binding site of MtrA differed greatly from all the other cobalamin-binding sites. Nevertheless, the hydrogen-bond linkage at the lower axial-ligand site of cobalt was equivalently constructed to that found in other methyltransferases and mutases. A distinct polypeptide segment fixed through the hydrogen-bond linkage in the relaxed Co(III) state might be involved in propagating the energy released upon corrinoid demethylation to the sodium-translocation site by a conformational change. PMID:27324530

MTR BUILDING INTERIOR, TRA603. BASEMENT. CAMERA IN WEST CORRIDOR FACING ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR BUILDING INTERIOR, TRA-603. BASEMENT. CAMERA IN WEST CORRIDOR FACING SOUTH. FREIGHT ELEVATOR IS AT RIGHT OF VIEW. AT CENTER VIEW IS MTR VAULT NO. 1, USED TO STORE SPECIAL OR FISSIONABLE MATERIALS. INL NEGATIVE NO. HD46-6-3. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durham, J. M.; Poulson, D.; Bacon, J.

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics providesmore » a potential solution to this long-standing problem in international nuclear safeguards.« less

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

DOE PAGES

Durham, J. M.; Poulson, D.; Bacon, J.; ...

2018-04-10

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics providesmore » a potential solution to this long-standing problem in international nuclear safeguards.« less

Next Generation Safeguards Initiative research to determine the Pu mass in spent fuel assemblies: Purpose, approach, constraints, implementation, and calibration

NASA Astrophysics Data System (ADS)

Tobin, S. J.; Menlove, H. O.; Swinhoe, M. T.; Schear, M. A.

2011-10-01

The Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy has funded a multi-lab/multi-university collaboration to quantify the plutonium mass in spent nuclear fuel assemblies and to detect the diversion of pins from them. The goal of this research effort is to quantify the capability of various non-destructive assay (NDA) technologies as well as to train a future generation of safeguards practitioners. This research is "technology driven" in the sense that we will quantify the capabilities of a wide range of safeguards technologies of interest to regulators and policy makers; a key benefit to this approach is that the techniques are being tested in a unified manner. When the results of the Monte Carlo modeling are evaluated and integrated, practical constraints are part of defining the potential context in which a given technology might be applied. This paper organizes the commercial spent fuel safeguard needs into four facility types in order to identify any constraints on the NDA system design. These four facility types are the following: future reprocessing plants, current reprocessing plants, once-through spent fuel repositories, and any other sites that store individual spent fuel assemblies (reactor sites are the most common facility type in this category). Dry storage is not of interest since individual assemblies are not accessible. This paper will overview the purpose and approach of the NGSI spent fuel effort and describe the constraints inherent in commercial fuel facilities. It will conclude by discussing implementation and calibration of measurement systems. This report will also provide some motivation for considering a couple of other safeguards concepts (base measurement and fingerprinting) that might meet the safeguards need but not require the determination of plutonium mass.

Spent fuel reaction - the behavior of the {epsilon}-phase over 3.1 years

DOE Office of Scientific and Technical Information (OSTI.GOV)

Finn, P.A.; Hoh, J.C.; Wolf, S.F.

The release fractions of the five elements in the {epsilon}-phase ({sup 99}Tc, {sup 97}Mo, Ru, Rh, and Pd) as well as that of {sup 238}U are reported for the reaction of two oxide fuels (ATM-103 and ATM-106) in unsaturated tests under oxidizing conditions. The {sup 99}Tc release fractions provide a lower limit for the magnitude of the spent fuel reaction. The {sup 99}Tc release fractions indicate that a surface reaction might be the rate controlling mechanism for fuel reaction under unsaturated conditions and the oxidant is possibly H{sub 2}O{sub 2}, a product of alpha radiolysis of water.

Separation of uranium from technetium in recovery of spent nuclear fuel

DOEpatents

Pruett, D.J.; McTaggart, D.R.

1983-08-31

Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc/sup +7/ therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

Separation of uranium from technetium in recovery of spent nuclear fuel

DOEpatents

Pruett, David J.; McTaggart, Donald R.

1984-01-01

Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc.sup.+7 therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR).

Container materials in environments of corroded spent nuclear fuel

NASA Astrophysics Data System (ADS)

Huang, F. H.

1996-07-01

Efforts to remove corroded uranium metal fuel from the K Basins wet storage to long-term dry storage are underway. The multi-canister overpack (MCO) is used to load spent nuclear fuel for vacuum drying, staging, and hot conditioning; it will be used for interim dry storage until final disposition options are developed. Drying and conditioning of the corroded fuel will minimize the possibility of gas pressurization and runaway oxidation. During all phases of operations the MCO is subjected to radiation, temperature and pressure excursions, hydrogen, potential pyrophoric hazard, and corrosive environments. Material selection for the MCO applications is clearly vital for safe and efficient long-term interim storage. Austenitic stainless steels (SS) such as 304L SS or 316L SS appear to be suitable for the MCO. Of the two, Type 304L SS is recommended because it possesses good resistance to chemical corrosion, hydrogen embrittlement, and radiation-induced corrosive species. In addition, the material has adequate strength and ductility to withstand pressure and impact loading so that the containment boundary of the container is maintained under accident conditions without releasing radioactive materials.

Container for reprocessing and permanent storage of spent nuclear fuel assemblies

DOEpatents

Forsberg, C.W.

1992-03-24

A single canister process container is described for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining its integrity at a temperature necessary to oxidize the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container. 10 figs.

MTR and In-vivo 1H-MRS studies on mouse brain with parkinson's disease

NASA Astrophysics Data System (ADS)

Yoon, Moon-Hyun; Kim, Hyeon-Jin; Chung, Jin-Yeung; Doo, Ah-Reum; Park, Hi-Joon; Kim, Seung-Nam; Choe, Bo-Young

2012-12-01

The aim of this study was to investigate whether the changes in the magnetization transfer ratio (MTR) histogram are related to specific characteristics of Parkinson's disease (PD) and to investigate whether the MTR histogram parameters are associated with neurochemical dysfunction by performing in vivo proton magnetic resonance spectroscopy (1H-MRS). MTR and in vivo 1H-MRS studies were performed on control mice (n = 10) and 1-methyl-1,2,3,6-tetrahydropyridine intoxicated mice (n = 10). All the MTR and in vivo 1H-MRS experiments were performed on a 9.4 T MRI/MRS system (Bruker Biospin, Germany) using a standard head coil. The protondensity fast spin echo (FSE) images and the T2-weighted spin echo (SE) images were acquired with no gap. Outer volume suppression (OVS), combined with the ultra-short echo-time stimulated echo acquisition mode (STEAM), was used for the localized in-vivo 1H-MRS. The quantitative analysis of metabolites was performed from the 1H spectra obtained in vivo on the striatum (ST) by using jMRUI (Lyon, France). The peak height of the MTR histograms in the PD model group was significantly lower than that in the control group (p < 0.05). The midbrain MTR values for volume were lower in the PD group than the control group(p < 0.05). The complex peak (Glx: glutamine+glutamate+ GABA)/creatine (Cr) ratio of the right ST in the PD group was significantly increased as compared to that of the control group. The present study revealed that the peak height of the MTR histogram was significantly decreased in the ST and substantia nigra, and a significant increase in the Gl x /Cr ratio was found in the ST of the PD group, as compared with that of the control group. These findings could reflect the early phase of neuronal dysfunction of neurotransmitters.

Measuring the Multiplication of Spent Fuel Assemblies – It’s easier than you think!

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobin, Stephen Joseph

This is a set of eight slides which advertise how easy it can be to measure the multiplication of a spent fuel assembly. A robust (fission chambers), rapid (under 15 minutes), direct (multiplication is measured, not photons from fission fragments) measurement of multiplication is possible.

ATR Spent Fuel Options Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connolly, Michael James; Bean, Thomas E.; Brower, Jeffrey O.

The Advanced Test Reactor (ATR) is a materials and fuels test nuclear reactor that performs irradiation services for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Naval Reactors, the National Nuclear Security Administration (NNSA), and other research programs. ATR achieved initial criticality in 1967 and is expected to operate in support of needed missions until the year 2050 or beyond. It is anticipated that ATR will generate approximately 105 spent nuclear fuel (SNF) elements per year through the year 2050. Idaho National Laboratory (INL) currently stores 2,008 ATR SNF elements in dry storage, 976 in wet storage,more » and expects to have 1,000 elements in wet storage before January 2017. A capability gap exists at INL for long-term (greater than the year 2050) management, in compliance with the Idaho Settlement Agreement (ISA), of ATR SNF until a monitored retrievable geological repository is open. INL has significant wet and dry storage capabilities that are owned by the DOE Office of Environmental Management (EM) and operated and managed by Fluor Idaho, which include the Idaho Nuclear Technology and Engineering Center’s (INTEC’s) CPP-666, CPP-749, and CPP-603. In addition, INL has other capabilities owned by DOE-NE and operated and managed by Battelle Energy Alliance, LLC (BEA), which are located at the Materials and Fuel Complex (MFC). Additional storage capabilities are located on the INL Site at the Naval Reactors Facility (NRF). Current INL SNF management planning, as defined in the Fluor Idaho contract, shows INTEC dry fuel storage, which is currently used for ATR SNF, will be nearly full after transfer of an additional 1,000 ATR SNF from wet storage. DOE-NE tasked BEA with identifying and analyzing options that have the potential to fulfill this capability gap. BEA assembled a team comprised of SNF management experts from Fluor Idaho, Savannah River Site (SRS), INL/BEA, and the MITRE Corp with an objective of developing and

Radiation induced corrosion of copper for spent nuclear fuel storage

NASA Astrophysics Data System (ADS)

Björkbacka, Åsa; Hosseinpour, Saman; Johnson, Magnus; Leygraf, Christofer; Jonsson, Mats

2013-11-01

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

Measurement and Analysis Plan for Investigation of Spent-Fuel Assay Using Lead Slowing-Down Spectroscopy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Leon E.; Haas, Derek A.; Gavron, Victor A.

2009-09-25

Under funding from the Department of Energy Office of Nuclear Energy’s Materials, Protection, Accounting, and Control for Transmutation (MPACT) program (formerly the Advanced Fuel Cycle Initiative Safeguards Campaign), Pacific Northwest National Laboratory (PNNL) and Los Alamos National Laboratory (LANL) are collaborating to study the viability of lead slowing-down spectroscopy (LSDS) for spent-fuel assay. Based on the results of previous simulation studies conducted by PNNL and LANL to estimate potential LSDS performance, a more comprehensive study of LSDS viability has been defined. That study includes benchmarking measurements, development and testing of key enabling instrumentation, and continued study of time-spectra analysis methods.more » This report satisfies the requirements for a PNNL/LANL deliverable that describes the objectives, plans and contributing organizations for a comprehensive three-year study of LSDS for spent-fuel assay. This deliverable was generated largely during the LSDS workshop held on August 25-26, 2009 at Rensselaer Polytechnic Institute (RPI). The workshop itself was a prominent milestone in the FY09 MPACT project and is also described within this report.« less

75 FR 49813 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1, Confirmation of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-16

... Storage Casks: MAGNASTOR System, Revision 1, Confirmation of Effective Date AGENCY: Nuclear Regulatory... spent fuel storage regulations at 10 CFR 72.214 to revise the MAGNASTOR System listing to include...

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, M.M.; Lau, L.K.; Schulz, T.L.

1993-12-14

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

Local negotiation on compensation siting of the spent nuclear fuel repository in Finland

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kojo, Matti

The aim of the paper is to analyse the local negotiation process between the Municipality of Eurajoki and the nuclear power company Teollisuuden Voima (TVO) and the nuclear waste management company Posiva Oy. The aim of the negotiations was to find an acceptable form of compensation for siting a spent nuclear fuel repository in Olkiluoto, Finland. The paper includes background information on the siting process in Finland, the local political setting in the Municipality of Eurajoki and a description of the negotiation process. The analysis of the negotiations on compensation is important for better understanding the progress of the Finnishmore » siting process. The paper describes the picture of the contest to host the spent nuclear fuel repository. It also provides more information on the relationship between the Municipality of Eurajoki and the power company TVO. The negotiations on compensation and the roles of various players in the negotiations have not been studied in detail because the minutes of the Vuojoki liaison group were not available before the decision of the Supreme Administrative Court in May 2006. (author)« less

Robotic Spent Fuel Monitoring – It is time to improve old approaches and old techniques!

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobin, Stephen Joseph; Dasari, Venkateswara Rao; Trellue, Holly Renee

This report describes various approaches and techniques associated with robotic spent fuel monitoring. The purpose of this description is to improve the quality of measured signatures, reduce the inspection burden on the IAEA, and to provide frequent verification.

PUMP HOUSE FOR MTR WELL NO. 1, TRA601. FLOOR PLAN, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PUMP HOUSE FOR MTR WELL NO. 1, TRA-601. FLOOR PLAN, ELEVATIONS, SECTION SHOWING WELL CASING, ROOF FRAMING PLAN. AS BUILT. WELL HOUSE FOR WELL NO. 2, TRA-602, WAS IDENTICAL IN ALL PARTICULARS EXCEPT FLOOR DIMENSIONS AND ARRANGEMENT OF PUMP AND ELECTRICAL EQUIPMENT INSIDE. IDAHO OPERATIONS OFFICE MTR-601-IDO-1, 12/1954. INL INDEX NO. 531-0601-00-396-110463, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

NEUTRON CHARACTERIZATION OF ENSA-DPT TYPE SPENT FUEL CASK AT TRILLO NUCLEAR POWER PLANT.

PubMed

Méndez-Villafañe, Roberto; Campo-Blanco, Xandra; Embid, Miguel; Yéboles, César A; Morales, Ramón; Novo, Manuel; Sanz, Javier

2018-04-23

The Neutron Standards Laboratory of CIEMAT has conducted the characterization of the independent spent fuel storage installation at the Trillo Nuclear Power Plant. At this facility, the spent fuel assemblies are stored in ENSA-DPT type dual purpose casks. Neutron characterization was performed by dosimetry measurements with a neutron survey meter (LB6411) inside the facility, around an individual cask and between stored casks, and outside the facility. Spectra measurements were also performed with a Bonner sphere system in order to determine the integral quantities and validate the use of the neutron monitor at the different positions. Inside the facility, measured neutron spectra and neutron ambient dose equivalent rate are consistent with the casks spatial distribution and neutron emission rates, and measurements with both instruments are consistent with each other. Outside the facility, measured neutron ambient dose equivalent rates are well below the 0.5 μSv/h limit established by the nuclear regulatory authority.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Peterson, Joshua L.; Gauld, Ian C.

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, decaymore » 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.« less

10 CFR 51.61 - Environmental report-independent spent fuel storage installation (ISFSI) or monitored retrievable...

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Environmental report-independent spent fuel storage installation (ISFSI) or monitored retrievable storage installation (MRS) license. 51.61 Section 51.61 Energy... amended at 68 FR 58811, Oct. 10, 2003] ...

Global Spent Fuel Logistics Systems Study (GSFLS). Volume 3A. GSFLS technical analysis (appendix). Interim report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kriger, A.

1978-01-31

This report is a part of the interim report documentation for the Global Spent Fuel Logistics System (GSFLS) study. The technical and financial considerations underlying a global spent fuel logistics systems have been studied and are reported. The Pacific Basin is used as a model throughout this report; however the stated methodology and, in many cases, considerations and conclusions are applicable to other global regions. Spent fuel discharge profiles for Pacific Basin Countries were used to determine the technical systems requirements for alternative concepts. Functional analyses and flows were generated to define both system design requirements and logistics parameters. Amore » technology review was made to ascertain the state-of-the-art of relevant GSFLS technical systems. Modular GSFLS facility designs were developed using the information generated from the functional analysis and technology review. The modular facility designs were used as a basis for siting and cost estimates for various GSFLS alternatives. Various GSFLS concepts were analyzed from a financial and economic perspective in order to provide total concepts costs and ascertain financial and economic sensitivities to key GSFLS variations. Results of the study include quantification of GSFLS facility and hardware requirements; drawings of relevant GSFLS facility designs; system cost estimates; financial reports - including user service charges; and comparative analyses of various GSFLS alternatives.« less

Sequestration of radioactive iodine in silver-palladium phases in commercial spent nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buck, Edgar C.; Mausolf, Edward J.; McNamara, Bruce K.

Radioactive iodine is the Achilles’ heel in the design for the safe geological disposal of spent UO2 nuclear fuel. Iodine’s high solubility and anticipated instant release during waste package compromise jeopardize performance assessment calculations. However, dissolution studies have indicated that the instant release fraction (IRF) of radioiodine (I) does not correlate with increasing fuel burn-up. In fact, there is a peak in the release iodine at around 50-60 Mwd/kgU and with increasing burn-up the instant release of iodine decreases. Detailed electron microscopy analysis of high burn-up fuel (~80 MWd/kgU) has revealed the presence of (Pd,Ag)(I,Br) nano-particles. As UO2 fuels aremore » irradiated, the Ag and Pd content increases, from 239Pu fission, enabling radioiodine to be retained. The occurrence of these phases in nuclear fuels may have significant implications for the long-term behavior of iodine.« less

Dehydrogenation of ammonia-borane by cationic Pd(II) and Ni(II) complexes in a nitromethane medium: hydrogen release and spent fuel characterization.

PubMed

Kim, Sung-Kwan; Hong, Sung-Ahn; Son, Ho-Jin; Han, Won-Sik; Michalak, Artur; Hwang, Son-Jong; Kang, Sang Ook

2015-04-28

A highly electrophilic cationic Pd(II) complex, [Pd(MeCN)4][BF4]2 (1), brings about the preferential activation of the B-H bond in ammonia-borane (NH3·BH3, AB). At room temperature, the reaction between 1 in CH3NO2 and AB in tetraglyme leads to Pd nanoparticles and formation of spent fuels of the general formula MeNHxBOy as reaction byproducts, while 2 equiv. of H2 is efficiently released per AB equiv. at room temperature within 60 seconds. For a mechanistic understanding of dehydrogenation by 1, the chemical structures of spent fuels were intensely characterized by a series of analyses such as elemental analysis (EA), X-ray photoelectron spectroscopy (XPS), solid state magic-angle-spinning (MAS) NMR spectra ((2)H, (13)C, (15)N, and (11)B), and cross polarization (CP) MAS methods. During AB dehydrogenation, the involvement of MeNO2 in the spent fuels showed that the mechanism of dehydrogenation catalyzed by 1 is different from that found in the previously reported results. This AB dehydrogenation derived from MeNO2 is supported by a subsequent digestion experiment of the AB spent fuel: B(OMe)3 and N-methylhydroxylamine ([Me(OH)N]2CH2), which are formed by the methanolysis of the AB spent fuel (MeNHxBOy), were identified by means of (11)B NMR and single crystal structural analysis, respectively. A similar catalytic behavior was also observed in the AB dehydrogenation catalyzed by a nickel catalyst, [Ni(MeCN)6][BF4]2 (2).

Research on Spent Fuel Storage and Transportation in CRIEPI (Part 2 Concrete Cask Storage)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koji Shirai; Jyunichi Tani; Taku Arai

2008-10-01

Concrete cask storage has been implemented in the world. At a later stage of storage period, the containment of the canister may deteriorate due to stress corrosion cracking phenomena in a salty air environment. High resistant stainless steels against SCC have been tested as compared with normal stainless steel. Taking account of the limited time-length of environment with certain level of humidity and temperature range, the high resistant stainless steels will survive from SCC damage. In addition, the adhesion of salt from salty environment on the canister surface will be further limited with respect to the canister temperature and anglemore » of the canister surface against the salty air flow in the concrete cask. Optional countermeasure against SCC with respect to salty air environment has been studied. Devices consisting of various water trays to trap salty particles from the salty air were designed to be attached at the air inlet for natural cooling of the cask storage building. Efficiency for trapping salty particles was evaluated. Inspection of canister surface was carried out using an optical camera inserted from the air outlet through the annulus of a concrete cask that has stored real spent fuel for more than 15 years. The camera image revealed no gross degradation on the surface of the canister. Seismic response of a full-scale concrete cask with simulated spent fuel assemblies has been demonstrated. The cask did not tip over, but laterally moved by the earthquake motion. Stress generated on the surface of the spent fuel assemblies during the earthquake motion were within the elastic region.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sindelar, R.; Deible, R.

2011-04-27

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

Direct measurement of 235U in spent fuel rods with Gamma-ray mirrors

NASA Astrophysics Data System (ADS)

Ruz, J.; Brejnholt, N. F.; Alameda, J. B.; Decker, T. A.; Descalle, M. A.; Fernandez-Perea, M.; Hill, R. M.; Kisner, R. A.; Melin, A. M.; Patton, B. W.; Soufli, R.; Ziock, K.; Pivovaroff, M. J.

2015-03-01

Direct measurement of plutonium and uranium X-rays and gamma-rays is a highly desirable non-destructive analysis method for the use in reprocessing fuel environments. The high background and intense radiation from spent fuel make direct measurements difficult to implement since the relatively low activity of uranium and plutonium is masked by the high activity from fission products. To overcome this problem, we make use of a grazing incidence optic to selectively reflect Kα and Kβ fluorescence of Special Nuclear Materials (SNM) into a high-purity position-sensitive germanium detector and obtain their relative ratios.

Determination of in situ state of stress at the Spent Fuel Test-Climax site, Climax Stock, Nevada Test Site

USGS Publications Warehouse

Ellis, W.L.; Magner, J.E.

1982-01-01

Determination of the in situ state of stress at the site of the Spent Fuel Test--Climax, using the U.S. Bureau of Mines overcore method, indicates principal stress magnitudes of 11.6 MPa, 7.1 MPa, and 2.8 MPa. The bearing and plunge of the maximum and minimum principal stress components are, respectively: N. 56? E., 29? NE; and N. 42? W., 14? NW. The vertical stress magnitude of 7.9 MPa calculated from the overcore data is significantly less than expected from overburden pressure, suggesting the stress field is influenced by local or areal geologic factors. Results from this investigation indicate (1) the stress state at the Spent Fuel Test--Climax site deviates significantly from a gravitational stress field, both in relative stress magnitudes and in orientation; (2) numerical modeling will not realistically simulate the near-field response of the Spent Fuel Test--Climax site if gravitational and (or) horizontal and vertical applied stress boundary conditions are assumed; and (3) substantial stress variations may occur spatially within the stock.

77 FR 60482 - Yankee Atomic Electric Company; Yankee Rowe Independent Spent Fuel Storage Installation, Staff...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-03

... Company; Yankee Rowe Independent Spent Fuel Storage Installation, Staff Evaluation; Exemption 1.0... exemption requests, the NRC staff believes that YAEC should be granted exemptions from the following.... Additional information regarding the NRC (staff) evaluation is documented in a Safety Evaluation Report that...

77 FR 48565 - Maine Yankee Atomic Power Company, Maine Yankee Independent Spent Fuel Storage Installation...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-14

... Power Company, Maine Yankee Independent Spent Fuel Storage Installation, Exemption--Staff Evaluation 1.0... in its November 29, 2010, letter. After evaluating the exemption requests, the staff determined that... staff evaluation is documented in a Safety Evaluation Report that contains Sensitive Unclassified Non...

77 FR 33005 - Connecticut Yankee Atomic Power Company; Haddam Neck Independent Spent Fuel Storage Installation...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-04

... Atomic Power Company; Haddam Neck Independent Spent Fuel Storage Installation, Staff Evaluation... requests, the staff determined CYAPCO should be granted exemptions from the following requirements: 10 CFR... regarding the NRC staff evaluation is documented in a Safety Evaluation Report that contains Sensitive...

Spent fuel canister for geological repository: Inner material requirements and candidates evaluation

NASA Astrophysics Data System (ADS)

Puig, Francesc; Dies, Javier; Pablo, Joan de; Martínez-Esparza, Aurora

2008-05-01

One of the key aspects in designing Spanish spent nuclear fuel canister for geological repository is selecting the inner material to be placed between the steel walls and the fuel assemblies. This material has to primarily avoid the possibility of a criticality event once the canister gets breached by corrosion and flooded by groundwater. A detailed set of requirements for a material to fulfil this role in that environment have been devised and presented in this paper. With these requirements in view, eight potentially interesting candidates were evaluated: cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, haematite, phosphates, and olivine. Among these, the first four materials or their families are found promising for this application.

Spent fuel measurements. passive neutron albedo reactivity (PNAR) and photon signatures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eigenbrodt, Julia; Menlove, Howard Olsen

2016-03-29

The International Atomic Energy Agency’s (IAEA) safeguards technical objective is the timely detection of a diversion of a significant quantity of nuclear material from peaceful activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection. An important IAEA task towards meeting this objective is the ability to accurately and reliably measure spent nuclear fuel (SNF) to verify reactor operating parameters and verify that the fuel has not been removed from reactors or SNF storage facilities. This dissertation analyzes a method to improvemore » the state-of-the-art of nuclear material safeguards measurements using two combined measurement techniques: passive neutron albedo reactivity (PNAR) and passive spectral photon measurements.« less

High Burn-Up Spent Nuclear Fuel Vibration Integrity Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong; Jiang, Hao

2015-01-01

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

Spent Fuel Assay with an Ultra-High Rate HPGe Spectrometer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fast, James; Fulsom, Bryan; Pitts, Karl

2015-07-01

Traditional verification of spent nuclear fuel (SNF) includes determination of initial enrichment, burnup and cool down time (IE, BU, CT). Along with neutron measurements, passive gamma assay provides important information for determining BU and CT. Other gamma-ray-based assay methods such as passive tomography and active delayed gamma offer the potential to measure the spatial distribution of fission products and the fissile isotopic concentration of the fuel, respectively. All fuel verification methods involving gamma-ray spectroscopy require that the spectrometers manage very high count rates while extracting the signatures of interest. PNNL has developed new digital filtering and analysis techniques to producemore » an ultra-high rate gamma-ray spectrometer from a standard coaxial high-purity germanium (HPGe) crystal. This 37% relative efficiency detector has been operated for SNF measurements at input count rates of 500-1300 kcps and throughput in excess of 150 kcps. Optimized filtering algorithms preserve the spectroscopic capability of the system even at these high rates. This paper will present the results of both passive and active SNF measurement performed with this system at PNNL. (authors)« less

TEST SYSTEM FOR EVALUATING SPENT NUCLEAR FUEL BENDING STIFFNESS AND VIBRATION INTEGRITY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong; Bevard, Bruce Balkcom

2013-01-01

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements specified by federal regulations. For normal conditions of transport, vibration loads incident to transport must be considered. This is particularly relevant for high-burnup fuel (>45 GWd/MTU). As the burnup of the fuel increases, a number of changes occur that may affect the performance of the fuel and cladding in storage and during transportation. The mechanical properties of high-burnup de-fueled cladding have been previously studied by subjecting defueled cladding tubes to longitudinal (axial) tensile tests, ring-stretch tests, ring-compression tests, and biaxial tube burst tests. The objective of this study ismore » to investigate the mechanical properties and behavior of both the cladding and the fuel in it under vibration/cyclic loads similar to the sustained vibration loads experienced during normal transport. The vibration loads to SNF rods during transportation can be characterized by dynamic, cyclic, bending loads. The transient vibration signals in a specified transport environment can be analyzed, and frequency, amplitude and phase components can be identified. The methodology being implemented is a novel approach to study the vibration integrity of actual SNF rod segments through testing and evaluating the fatigue performance of SNF rods at defined frequencies. Oak Ridge National Laboratory (ORNL) has developed a bending fatigue system to evaluate the response of the SNF rods to vibration loads. A three-point deflection measurement technique using linear variable differential transformers is used to characterize the bending rod curvature, and electromagnetic force linear motors are used as the driving system for mechanical loading. ORNL plans to use the test system in a hot cell for SNF vibration testing on high burnup, irradiated fuel to evaluate the pellet-clad interaction and bonding on the effective lifetime of fuel-clad structure bending fatigue performance

Helicase-Dependent RNA Decay Illuminated by a Cryo-EM Structure of a Human Nuclear RNA Exosome-MTR4 Complex.

PubMed

Weick, Eva-Maria; Puno, M Rhyan; Januszyk, Kurt; Zinder, John C; DiMattia, Michael A; Lima, Christopher D

2018-06-14

The ribonucleolytic RNA exosome interacts with RNA helicases to degrade RNA. To understand how the 3' to 5' Mtr4 helicase engages RNA and the nuclear exosome, we reconstituted 14-subunit Mtr4-containing RNA exosomes from Saccharomyces cerevisiae, Schizosaccharomyces pombe, and human and show that they unwind structured substrates to promote degradation. We loaded a human exosome with an optimized DNA-RNA chimera that stalls MTR4 during unwinding and determined its structure to an overall resolution of 3.45 Å by cryoelectron microscopy (cryo-EM). The structure reveals an RNA-engaged helicase atop the non-catalytic core, with RNA captured within the central channel and DIS3 exoribonuclease active site. MPP6 tethers MTR4 to the exosome through contacts to the RecA domains of MTR4. EXOSC10 remains bound to the core, but its catalytic module and cofactor C1D are displaced by RNA-engaged MTR4. Competition for the exosome core may ensure that RNA is committed to degradation by DIS3 when engaged by MTR4. Copyright © 2018 Elsevier Inc. All rights reserved.

MTR FAST NEUTRON FLUX MEASUREMENTS FOR CYCLE 146

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weber, L D; Hogg, C H

1962-03-20

The fast neutron fluxes in selected positions of the MTR were measured for Cycle 146. The measurements were made at the beginning, throughout, and at the end of the cycle (564 Mwd). Vertical traverses for each position monitors are shown. (auth)

A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

DOE PAGES

Tobin, Stephen J.; Lundkvist, Niklas; Goodsell, Alison V.; ...

2015-12-01

In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is bemore » ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.« less

A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobin, Stephen J.; Lundkvist, Niklas; Goodsell, Alison V.

In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is bemore » ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.« less

Processing and microstructural characterisation of a UO2-based ceramic for disposal studies on spent AGR fuel

NASA Astrophysics Data System (ADS)

Hiezl, Z.; Hambley, D. I.; Padovani, C.; Lee, W. E.

2015-01-01

Preparation and characterisation of a Simulated Spent Nuclear Fuel (SIMFuel), which replicates the chemical state and microstructure of Spent Nuclear Fuel (SNF) discharged from a UK Advanced Gas-cooled Reactor (AGR) after a cooling time of 100 years is described. Given the relatively small differences in radionuclide inventory expected over longer time periods, the SIMFuel studied in this work is expected to be also representative of spent fuel after significantly longer periods (e.g. 1000 years). Thirteen stable elements were added to depleted UO2 and sintered to simulate the composition of fuel pellets after burn-ups of 25 and 43 GWd/tU and, as a reference, pure UO2 pellets were also investigated. The fission product distribution was calculated using the FISPIN code provided by the UK National Nuclear Laboratory. SIMFuel pellets were up to 92% dense and during the sintering process in H2 atmosphere Mo-Ru-Rh-Pd metallic precipitates and grey-phase ((Ba, Sr)(Zr, RE) O3 oxide precipitates) formed within the UO2 matrix. These secondary phases are present in real PWR and AGR SNF. Metallic precipitates are generally spherical and have submicron particle size (0.8 ± 0.7 μm). Spherical oxide precipitates in SIMFuel measured up to 30 μm in diameter, but no data were available in the public domain to compare this to AGR SNF. The grain size of actual AGR SNF (∼ 3-30 μm) is larger than that measured in AGR SIMFuel (∼ 2-5 μm).

77 FR 24585 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-25

... Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear Regulatory Commission. ACTION: Direct final rule... revising the Holtec International HI-STORM 100 System listing within the ``List of Approved Spent Fuel...) 72.214, by revising the Holtec International HI-STORM 100 System listing within the ``List of...

Ultrasonic Fingerprinting of Structural Materials: Spent Nuclear Fuel Containers Case-Study

NASA Astrophysics Data System (ADS)

Sednev, D.; Lider, A.; Demyanuk, D.; Kroening, M.; Salchak, Y.

Nowadays, NDT is mainly focused on safety purposes, but it seems possible to apply those methods to provide national and IAEA safeguards. The containment of spent fuel in storage casks could be dramatically improved in case of development of so-called "smart" spent fuel storage and transfer casks. Such casks would have tamper indicating and monitoring/tracking features integrated directly into the cask design. The microstructure of the containers material as well as of the dedicated weld seam is applied to the lid and the cask body and provides a unique fingerprint of the full container, which can be reproducibly scanned by using an appropriate technique. The echo-sounder technique, which is the most commonly used method for material inspection, was chosen for this project. The main measuring parameter is acoustic noise, reflected from material's artefacts. The purpose is to obtain structural fingerprinting. Reference measurement and additional measurement results were compared. Obtained results have verified the appliance of structural fingerprint and the chosen control method. The successful authentication demonstrates the levels of the feature points' compliance exceeding the given threshold which differs considerably from the percentage of the concurrent points during authentication from other points. Since reproduction or doubling of the proposed unique identification characteristics is impossible at the current state science and technology, application of this technique is considered to identify the interference into the nuclear materials displacement with high accuracy.

Characterization of hMTr1, a Human Cap1 2′-O-Ribose Methyltransferase*

PubMed Central

Bélanger, François; Stepinski, Janusz; Darzynkiewicz, Edward; Pelletier, Jerry

2010-01-01

Cellular eukaryotic mRNAs are capped at their 5′ ends with a 7-methylguanosine nucleotide, a structural feature that has been shown to be important for conferring mRNA stability, stimulating mRNA biogenesis (splicing, poly(A) addition, nucleocytoplasmic transport), and increasing translational efficiency. Whereas yeast mRNAs have no additional modifications to the cap, called cap0, higher eukaryotes are methylated at the 2′-O-ribose of the first or the first and second transcribed nucleotides, called cap1 and cap2, respectively. In the present study, we identify the methyltransferase responsible for cap1 formation in human cells, which we call hMTr1 (also known as FTSJD2 and ISG95). We show in vitro that hMTr1 catalyzes specific methylation of the 2′-O-ribose of the first nucleotide of a capped RNA transcript. Using siRNA-mediated knockdown of hMTr1 in HeLa cells, we demonstrate that hMTr1 is responsible for cap1 formation in vivo. PMID:20713356

Life Prediction of Spent Fuel Storage Canister Material

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ballinger, Ronald

The original purpose of this project was to develop a probabilistic model for SCC-induced failure of spent fuel storage canisters, exposed to a salt-air environment in the temperature range 30-70°C for periods up to and exceeding 100 years. The nature of this degradation process, which involves multiple degradation mechanisms, combined with variable and uncertain environmental conditions dictates a probabilistic approach to life prediction. A final report for the original portion of the project was submitted earlier. However, residual stress measurements for as-welded and repair welds could not be performed within the original time of the project. As a result ofmore » this, a no-cost extension was granted in order to complete these tests. In this report, we report on the results of residual stress measurements.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Head, C.R.

1997-08-01

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

Casting technology for manufacturing metal rods from simulated metallic spent fuels

NASA Astrophysics Data System (ADS)

Leeand, Y. S.; Lee, D. B.; Kim, C. K.; Shin, Y. J.; Lee, J. H.

2000-09-01

A uranium metal rod 13.5 mm in diameter and 1,150 mm long was produced from simulated metallic spent fuels with advanced casting equipment using the directional-solidification method. A vacuum casting furnace equipped with a four-zone heater to prevent surface oxidation and the formation of surface shrinkage holes was designed. By controlling the axial temperature gradient of the casting furnace, deformation by the surface shrinkage phenomena was diminished, and a sound rod was manufactured. The cooling behavior of the molten uranium was analyzed using the computer software package MAGMAsoft.

Bread: CDC 7600 program that processes Spent Fuel Test Climax data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hage, G.L.

BREAD will process a family of files copied from a data tape made by Hewlett-Packard equipment employed for data acquisition on the Spent Fuel Test-Climax at NTS. Tapes are delivered to Livermore approximately monthly. The process at this stage consists of four steps: read the binary files and convert from H-P 16-bit words to CDC 7600 60-bit words; check identification and data ranges; write the data in 6-bit ASCII (BCD) format, one data point per line; then sort the file by identifier and time.

10 CFR 51.61 - Environmental report-independent spent fuel storage installation (ISFSI) or monitored retrievable...

Code of Federal Regulations, 2011 CFR

2011-01-01

... either an environmental impact statement or an environmental assessment, as appropriate. The... factors contained in subpart E of part 72 of this chapter. Unless otherwise required by the Commission, in... the environmental impact of the storage of spent fuel at an ISFSI beyond the term of the license or...

Structure and function of Neisseria gonorrhoeae MtrF illuminates a class of antimetabolite efflux pumps

DOE PAGES

Su, Chih -Chia; Bolla, Jani  Reddy; Kumar, Nitin; ...

2015-04-01

Neisseria gonorrhoeae is an obligate human pathogen and the causative agent of the sexually transmitted disease gonorrhea. The control of this disease has been compromised by the increasing proportion of infections due to antibiotic-resistant strains, which are growing at an alarming rate. N. gonorrhoeae MtrF is an integral membrane protein that belongs to the AbgT family of transporters for which no structural information is available. Here, we describe the crystal structure of MtrF, revealing a dimeric molecule with architecture distinct from all other families of transporters. MtrF is a bowl-shaped dimer with a solvent-filled basin extending from the cytoplasm tomore » halfway across the membrane bilayer. Each subunit of the transporter contains nine transmembrane helices and two hairpins, posing a plausible pathway for substrate transport. A combination of the crystal structure and biochemical functional assays suggests that MtrF is an antibiotic efflux pump mediating bacterial resistance to sulfonamide antimetabolite drugs.« less

Structure and function of Neisseria gonorrhoeae MtrF illuminates a class of antimetabolite efflux pumps

PubMed Central

Su, Chih-Chia; Bolla, Jani Reddy; Kumar, Nitin; Radhakrishnan, Abhijith; Long, Feng; Delmar, Jared A.; Chou, Tsung-Han; Rajashankar, Kanagalaghatta R.; Shafer, William M.; Yu, Edward W.

2015-01-01

SUMMARY Neisseria gonorrhoeae is an obligate human pathogen and the causative agent of the sexually-transmitted disease gonorrhea. The control of this disease has been compromised by the increasing proportion of infections due to antibiotic-resistant strains, which are growing at an alarming rate. N. gonorrhoeae MtrF is an integral membrane protein, which belongs to the AbgT family of transporters for which no structural information is available. Here we describe the crystal structure of MtrF, revealing a dimeric molecule with architecture distinct from all other families of transporters. MtrF is a bowl-shaped dimer with a solvent-filled basin extending from the cytoplasm to halfway across the membrane bilayer. Each subunit of the transporter contains nine transmembrane helices and two hairpins, posing a plausible pathway for substrate transport. A combination of the crystal structure and biochemical functional assays suggests that MtrF is an antibiotic efflux pump, mediating bacterial resistance to sulfonamide antimetabolite drugs. PMID:25818299

The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

DOE PAGES

Jiang, Hao; Wang, Jy-An John; Wang, Hong

2016-09-26

Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets tomore » the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.« less

The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, Hao; Wang, Jy-An John; Wang, Hong

Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets tomore » the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.« less

Achievements and Perspectives of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Louvat, D.; Lacoste, A.C.

The Joint Convention on the Safety of Spent Fuel management and on the Safety of Radioactive Waste Management is the first legal instrument to directly address the safety of spent fuel and radioactive waste management on a global scale. The Joint Convention entered into force in 2001. This paper describes its process and its main achievements to date. The perspectives to establish of a Global Waste Safety Regime based on the Joint Convention are also discussed. (authors)

Radiolytic and thermal process relevant to dry storage of spent nuclear fuels. 1998 annual progress report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marschman, S.C.; Cowin, J.P.; Orlando, T.M.

1998-06-01

'This project involves basic research in chemistry and physics aimed at providing information pertinent to the safe long-term dry storage of spent nuclear fuel (SNF), thousands of tons of which remain in water storage across the DOE complex. The Hanford Site K-Basins alone hold 2,300 tons of spent fuel, much of it severely corroded, and similar situations exist at Savannah River and Idaho National Engineering and Environmental Laboratory. The DOE plans to remove this fuel and seal it in overpack canisters for dry interim storage for up to 75 years while awaiting permanent disposition. Chemically-bound water will remain in thismore » fuel even following proposed drying steps, leading to possible long-term corrosion of the containers and/or fuel rods themselves, generation of H{sub 2} and O{sub 2} gas via radiolysis (which could lead to deflagration or detonation), and reactions of pyrophoric uranium hydrides. No thoroughly tested model is currently available to predict fuel behavior during pre-processing, processing, or storage. In a collaboration between Rutgers University, Pacific Northwest National Laboratory, and Brookhaven National Laboratory, the authors are studying the radiolytic reaction, drying processes, and corrosion behavior of actual SNF materials, and of pure and mixed-phase samples. The authors propose to determine what is omitted from current models: radiolysis of water adsorbed on or in hydrates or hydroxides, thermodynamics of interfacial phases, and kinetics of drying. A model will be developed and tested against actual fuel rod behavior to insure validity and applicability to the problems associated with developing dry storage strategies for DOE-owned SNF. This report summarizes work after eight months of a three-year project.'« less

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

DOE PAGES

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael; ...

2016-11-17

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

Programming the quorum sensing-based AND gate in Shewanella oneidensis for logic gated-microbial fuel cells.

PubMed

Hu, Yidan; Yang, Yun; Katz, Evgeny; Song, Hao

2015-03-11

An AND logic gate based on a synthetic quorum-sensing (QS) module was constructed in a Shewanella oneidensis MR-1 mtrA knockout mutant. The presence of two input signals activated the expression of a periplasmic decaheme cytochrome MtrA to regenerate the extracellular electron transfer conduit, enabling the construction of AND-gated microbial fuel cells.

Calculation of the process of vacuum drying of a metal-concrete container with spent nuclear fuel

NASA Astrophysics Data System (ADS)

Karyakin, Yu. E.; Lavrent'ev, S. A.; Pavlyukevich, N. V.; Pletnev, A. A.; Fedorovich, E. D.

2012-01-01

An algorithm and results of calculation of the process of vacuum drying of a metal-concrete container intended for long-term "dry" storage of spent nuclear fuel are presented. A calculated substantiation of the initial amount of moisture in the container is given.

Role of Mex67-Mtr2 in the Nuclear Export of 40S Pre-Ribosomes

PubMed Central

Occhipinti, Laura; Kemmler, Stefan; Panse, Vikram G.

2012-01-01

Nuclear export of mRNAs and pre-ribosomal subunits (pre40S and pre60S) is fundamental to all eukaryotes. While genetic approaches in budding yeast have identified bona fide export factors for mRNAs and pre60S subunits, little is known regarding nuclear export of pre40S subunits. The yeast heterodimeric transport receptor Mex67-Mtr2 (TAP-p15 in humans) binds mRNAs and pre60S subunits in the nucleus and facilitates their passage through the nuclear pore complex (NPC) into the cytoplasm by interacting with Phe-Gly (FG)-rich nucleoporins that line its transport channel. By exploiting a combination of genetic, cell-biological, and biochemical approaches, we uncovered an unanticipated role of Mex67-Mtr2 in the nuclear export of 40S pre-ribosomes. We show that recruitment of Mex67-Mtr2 to pre40S subunits requires loops emanating from its NTF2-like domains and that the C-terminal FG-rich nucleoporin interacting UBA-like domain within Mex67 contributes to the transport of pre40S subunits to the cytoplasm. Remarkably, the same loops also recruit Mex67-Mtr2 to pre60S subunits and to the Nup84 complex, the respective interactions crucial for nuclear export of pre60S subunits and mRNAs. Thus Mex67-Mtr2 is a unique transport receptor that employs a common interaction surface to participate in the nuclear export of both pre-ribosomal subunits and mRNAs. Mex67-Mtr2 could engage a regulatory crosstalk among the three major export pathways for optimal cellular growth and proliferation. PMID:22956913

MTR WING, TRA604, INTERIOR. BASEMENT. INTERIOR VIEW FROM SAME LOCATION ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR WING, TRA-604, INTERIOR. BASEMENT. INTERIOR VIEW FROM SAME LOCATION IN WEST CORRIDOR AS PHOTO ID-33-G-42 BUT CAMERA FACES SOUTH. SIGN ON DOOR FOR "PIPE TUNNEL" WARNS OF RADIOLOGICAL AND ASBESTOS HAZARDS. DOOR HAS METAL HASPS. SIGN ON OVERHEAD WASTE HEAT RECOVERY PIPES SAYS THEY CONTAIN "ASBESTOS FREE INSULATION." FIRE DOOR AT LEFT LEADS TO STAIRWAY TO FIRST FLOOR. DOOR AT RIGHT LEADS TO ROOM WHICH ONCE CONTAINED MTR LIBRARY. INL NEGATIVE NO. HD46-13-4. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Measuring spent fuel assembly multiplication in borated water with a passive neutron albedo reactivity instrument

NASA Astrophysics Data System (ADS)

Tobin, Stephen J.; Peura, Pauli; Bélanger-Champagne, Camille; Moring, Mikael; Dendooven, Peter; Honkamaa, Tapani

2018-07-01

The performance of a passive neutron albedo reactivity (PNAR) instrument to measure neutron multiplication of spent nuclear fuel in borated water is investigated as part of an integrated non-destructive assay safeguards system. To measure the PNAR Ratio, which is proportional to the neutron multiplication, the total neutron count rate is measured in high- and low-multiplying environments by the PNAR instrument. The integrated system also contains a load cell and a passive gamma emission tomograph, and as such meets all the recommendations of the IAEA's recent ASTOR Experts Group report. A virtual spent fuel library for VVER-440 fuel was used in conjunction with MCNP simulations of the PNAR instrument to estimate the measurement uncertainties from (1) variation in the water boron content, (2) assembly positioning in the detector and (3) counting statistics. The estimated aggregate measurement uncertainty on the PNAR Ratio measurement is 0.008, to put this uncertainty in context, the difference in the PNAR Ratio between a fully irradiated assembly and this same assembly when fissile isotopes only absorb neutrons, but do not emit neutrons, is 0.106, a 13-sigma effect. The 1-sigma variation of 0.008 in the PNAR Ratio is estimated to correspond to a 3.2 GWd/tU change in assembly burnup.

78 FR 78165 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-26

... Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9 AGENCY: Nuclear Regulatory... storage regulations by revising the Holtec International HI-STORM 100 Cask System listing within the...

Mutation in a locus linked to penB-nmp causes suppression of the Mtr phenotype of Neisseria gonorrhoeae.

PubMed Central

Shinners, E N; Catlin, B W

1988-01-01

The chromosomal locus mtr, which encodes low-level resistance to multiple antibacterial agents in Neisseria gonorrhoeae, is subject to phenotypic suppression by env mutations that increase the permeability of the envelope. We have identified a new locus, mom (for modifier of Mtr), which is located on the chromosome very close to penB and nmp, loci known to be linked to each other and to spc. Phenotypic suppression of Mtr was recognized by reductions of resistance to benzylpenicillin and also to oxacillin and the hydrophobic agents novobiocin and erythromycin. The resistance to each of these antibiotics returned to the Mtr levels in mom+ transformants isolated by selection for increased resistance to either novobiocin or erythromycin; the accompanying change of the outer membrane protein I seroreactions confirmed the proximity of nmp and mom. Thus, some mutant gonococci display wild-type antibiotic susceptibilities but can express multiple resistance following a mom+ mutation that releases the suppressed Mtr phenotype. PMID:3142343

A multiphase interfacial model for the dissolution of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Jerden, James L.; Frey, Kurt; Ebert, William

2015-07-01

The Fuel Matrix Dissolution Model (FMDM) is an electrochemical reaction/diffusion model for the dissolution of spent uranium oxide fuel. The model was developed to provide radionuclide source terms for use in performance assessment calculations for various types of geologic repositories. It is based on mixed potential theory and consists of a two-phase fuel surface made up of UO2 and a noble metal bearing fission product phase in contact with groundwater. The corrosion potential at the surface of the dissolving fuel is calculated by balancing cathodic and anodic reactions occurring at the solution interfaces with UO2 and NMP surfaces. Dissolved oxygen and hydrogen peroxide generated by radiolysis of the groundwater are the major oxidizing agents that promote fuel dissolution. Several reactions occurring on noble metal alloy surfaces are electrically coupled to the UO2 and can catalyze or inhibit oxidative dissolution of the fuel. The most important of these is the oxidation of hydrogen, which counteracts the effects of oxidants (primarily H2O2 and O2). Inclusion of this reaction greatly decreases the oxidation of U(IV) and slows fuel dissolution significantly. In addition to radiolytic hydrogen, large quantities of hydrogen can be produced by the anoxic corrosion of steel structures within and near the fuel waste package. The model accurately predicts key experimental trends seen in literature data, the most important being the dramatic depression of the fuel dissolution rate by the presence of dissolved hydrogen at even relatively low concentrations (e.g., less than 1 mM). This hydrogen effect counteracts oxidation reactions and can limit fuel degradation to chemical dissolution, which results in radionuclide source term values that are four or five orders of magnitude lower than when oxidative dissolution processes are operative. This paper presents the scientific basis of the model, the approach for modeling used fuel in a disposal system, and preliminary

75 FR 81031 - Consideration of Environmental Impacts of Temporary Storage of Spent Fuel After Cessation of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-23

... Part VI Nuclear Regulatory Commission 10 CFR Part 51 Consideration of Environmental Impacts of... Consideration of Environmental Impacts of Temporary Storage of Spent Fuel After Cessation of Reactor Operation... Commission (NRC or Commission) is revising its generic determination on the environmental impacts of storage...

Testing and COBRA-SFS analysis of the VSC-17 ventilated concrete, spent fuel storage cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

McKinnon, M.A.; Dodge, R.E.; Schmitt, R.C.

1992-04-01

A performance test of a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask loaded with 17 canisters of consolidated PWR spent fuel generating approximately 15 kW was conducted. The performance test included measuring the cask surface, concrete, air channel surface, and fuel temperatures, as well as cask surface gamma and neutron dose rates. Testing was performed using vacuum, nitrogen, and helium backfill environments. Pretest predictions of cask thermal performance were made using the COBRA-SFS computer code. Analysis results were within 15{degrees}C of measured peak fuel temperature. Peak fuel temperature for normal operation was 321{degrees}C. In general, the surface dose ratesmore » were less than 30 mrem/h on the side of the cask and 40 mrem/h on the top of the cask.« less

76 FR 17019 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-28

... Storage Casks: HI-STORM Flood/Wind Addition AGENCY: Nuclear Regulatory Commission. ACTION: Direct final... regulations to add the HI-STORM Flood/Wind cask system to the ``List of Approved Spent Fuel Storage Casks... cask designs. Discussion This rule will add the Holtec HI-STORM Flood/Wind (FW) cask system to the list...

76 FR 33121 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-08

... Storage Casks: HI-STORM Flood/Wind Addition AGENCY: Nuclear Regulatory Commission. ACTION: Direct final... regulations to add the Holtec HI-STORM Flood/Wind cask system to the ``List of Approved Spent Fuel Storage... Title 10 of the Code of Federal Regulations Section 72.214 to add the Holtec HI- STORM Flood/Wind cask...

Ageing management program for the Spanish low and intermediate level waste disposal and spent fuel and high-level waste centralised storage facilities

NASA Astrophysics Data System (ADS)

Zuloaga, P.; Ordoñez, M.; Andrade, C.; Castellote, M.

2011-04-01

The generic design of the centralised spent fuel storage facility was approved by the Spanish Safety Authority in 2006. The planned operational life is 60 years, while the design service life is 100 years. Durability studies and surveillance of the behaviour have been considered from the initial design steps, taking into account the accessibility limitations and temperatures involved. The paper presents an overview of the ageing management program set in support of the Performance Assessment and Safety Review of El Cabril low and intermediate level waste (LILW) disposal facility. Based on the experience gained for LILW, ENRESA has developed a preliminary definition of the Ageing Management Plan for the Centralised Interim Storage Facility of spent Fuel and High Level Waste (HLW), which addresses the behaviour of spent fuel, its retrievability, the confinement system and the reinforced concrete structure. It includes tests plans and surveillance design considerations, based on the El Cabril LILW disposal facility.

Actinides in metallic waste from electrometallurgical treatment of spent nuclear fuel

NASA Astrophysics Data System (ADS)

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

2003-09-01

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

MELCOR Model of the Spent Fuel Pool of Fukushima Dai-ichi Unit 4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carbajo, Juan J

2012-01-01

Unit 4 of the Fukushima Dai-ichi Nuclear Power Plant suffered a hydrogen explosion at 6:00 am on March 15, 2011, exactly 3.64 days after the earthquake hit the plant and the off-site power was lost. The earthquake occurred on March 11 at 2:47 pm. Since the reactor of this Unit 4 was defueled on November 29, 2010, and all its fuel was stored in the spent fuel pool (SFP4), it was first believed that the explosion was caused by hydrogen generated by the spent fuel, in particular, by the recently discharged core. The hypothetical scenario was: power was lost, coolingmore » to the SFP4 water was lost, pool water heated/boiled, water level decreased, fuel was uncovered, hot Zircaloy reacted with steam, hydrogen was generated and accumulated above the pool, and the explosion occurred. Recent analyses of the radioisotopes present in the water of the SFP4 and underwater video indicated that this scenario did not occur - the fuel in this pool was not damaged and was never uncovered the hydrogen of the explosion was apparently generated in Unit 3 and transported through exhaust ducts that shared the same chimney with Unit 4. This paper will try to answer the following questions: Could that hypothetical scenario in the SFP4 had occurred? Could the spent fuel in the SPF4 generate enough hydrogen to produce the explosion that occurred 3.64 days after the earthquake? Given the magnitude of the explosion, it was estimated that at least 150 kg of hydrogen had to be generated. As part of the investigations of this accident, MELCOR models of the SFP4 were prepared and a series of calculations were completed. The latest version of MELCOR, version 2.1 (Ref. 1), was employed in these calculations. The spent fuel pool option for BWR fuel was selected in MELCOR. The MELCOR model of the SFP4 consists of a total of 1535 fuel assemblies out of which 548 assemblies are from the core defueled on Nov. 29, 2010, 783 assemblies are older assemblies, and 204 are new/fresh assemblies. The total

Amplicon Sequencing Reveals Microbiological Signatures in Spent Nuclear Fuel Storage Basins

DOE PAGES

Bagwell, Christopher E.; Noble, Peter A.; Milliken, Charles E.; ...

2018-03-09

Water quality is an important determinant for the structural integrity of alloy cladded fuels and assemblies during long-term wet storage. Detailed characterization of a water filled storage basin for spent nuclear reactor fuel was performed following the formation and proliferation of an amorphous white flocculent. White precipitant was sampled throughout the storage basin for chemical and spectroscopic characterization, and environmental DNA was extracted for 454 pyrosequencing of bacterial 16S rRNA gene diversity. Accordingly, spectroscopic analyses indicated the precipitant to be primarily amorphous to crystalline aluminum (oxy) hydroxides with minor associated elemental components including Fe, Si, Ti, and U. High levelsmore » of organic carbon were co-localized with the precipitant relative to bulk dissolved organic concentrations. Bacterial densities were highly variable between sampling locations and with depth within the water filled storage basin; cell numbers ranged from 4 × 10 3to 4 × 104 cells/mL. Bacterial diversity that was physically associated with the aluminum (oxy) hydroxide complexes exceeded an estimated 4,000 OTUs/amplicon library (3% cutoff) and the majority of sequences were aligned to the families Burkholderiaceae (23%), Nitrospiraceae (23%), Hyphomicrobiaceae (17%), and Comamonadaceae (6%). We surmise that episodic changes in the physical and chemical properties of the basin contribute to the polymerization of aluminum (oxy) hydroxides, which in turn can chemisorb nutrients, carbon ligands and bacterial cells from the surrounding bulk aqueous phase. As such, these precipitants should establish favorable microhabitats for bacterial colonization and growth. Comparative analyses of 16S rRNA gene amplicon libraries across a selection of natural and engineered aquatic ecosystems were performed and microbial community and taxonomic signatures unique to the spent nuclear fuel (SNF) storage basin environment were revealed. These insights could

Amplicon Sequencing Reveals Microbiological Signatures in Spent Nuclear Fuel Storage Basins.

PubMed

Bagwell, Christopher E; Noble, Peter A; Milliken, Charles E; Li, Dien; Kaplan, Daniel I

2018-01-01

Water quality is an important determinant for the structural integrity of alloy cladded fuels and assemblies during long-term wet storage. Detailed characterization of a water filled storage basin for spent nuclear reactor fuel was performed following the formation and proliferation of an amorphous white flocculent. White precipitant was sampled throughout the storage basin for chemical and spectroscopic characterization, and environmental DNA was extracted for 454 pyrosequencing of bacterial 16S rRNA gene diversity. Accordingly, spectroscopic analyses indicated the precipitant to be primarily amorphous to crystalline aluminum (oxy) hydroxides with minor associated elemental components including Fe, Si, Ti, and U. High levels of organic carbon were co-localized with the precipitant relative to bulk dissolved organic concentrations. Bacterial densities were highly variable between sampling locations and with depth within the water filled storage basin; cell numbers ranged from 4 × 10 3 to 4 × 10 4 cells/mL. Bacterial diversity that was physically associated with the aluminum (oxy) hydroxide complexes exceeded an estimated 4,000 OTUs/amplicon library (3% cutoff) and the majority of sequences were aligned to the families Burkholderiaceae (23%), Nitrospiraceae (23%), Hyphomicrobiaceae (17%), and Comamonadaceae (6%). We surmise that episodic changes in the physical and chemical properties of the basin contribute to the polymerization of aluminum (oxy) hydroxides, which in turn can chemisorb nutrients, carbon ligands and bacterial cells from the surrounding bulk aqueous phase. As such, these precipitants should establish favorable microhabitats for bacterial colonization and growth. Comparative analyses of 16S rRNA gene amplicon libraries across a selection of natural and engineered aquatic ecosystems were performed and microbial community and taxonomic signatures unique to the spent nuclear fuel (SNF) storage basin environment were revealed. These insights could spur

Amplicon Sequencing Reveals Microbiological Signatures in Spent Nuclear Fuel Storage Basins

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bagwell, Christopher E.; Noble, Peter A.; Milliken, Charles E.

Water quality is an important determinant for the structural integrity of alloy cladded fuels and assemblies during long-term wet storage. Detailed characterization of a water filled storage basin for spent nuclear reactor fuel was performed following the formation and proliferation of an amorphous white flocculent. White precipitant was sampled throughout the storage basin for chemical and spectroscopic characterization, and environmental DNA was extracted for 454 pyrosequencing of bacterial 16S rRNA gene diversity. Accordingly, spectroscopic analyses indicated the precipitant to be primarily amorphous to crystalline aluminum (oxy) hydroxides with minor associated elemental components including Fe, Si, Ti, and U. High levelsmore » of organic carbon were co-localized with the precipitant relative to bulk dissolved organic concentrations. Bacterial densities were highly variable between sampling locations and with depth within the water filled storage basin; cell numbers ranged from 4 × 10 3to 4 × 104 cells/mL. Bacterial diversity that was physically associated with the aluminum (oxy) hydroxide complexes exceeded an estimated 4,000 OTUs/amplicon library (3% cutoff) and the majority of sequences were aligned to the families Burkholderiaceae (23%), Nitrospiraceae (23%), Hyphomicrobiaceae (17%), and Comamonadaceae (6%). We surmise that episodic changes in the physical and chemical properties of the basin contribute to the polymerization of aluminum (oxy) hydroxides, which in turn can chemisorb nutrients, carbon ligands and bacterial cells from the surrounding bulk aqueous phase. As such, these precipitants should establish favorable microhabitats for bacterial colonization and growth. Comparative analyses of 16S rRNA gene amplicon libraries across a selection of natural and engineered aquatic ecosystems were performed and microbial community and taxonomic signatures unique to the spent nuclear fuel (SNF) storage basin environment were revealed. These insights could

Amplicon Sequencing Reveals Microbiological Signatures in Spent Nuclear Fuel Storage Basins

PubMed Central

Bagwell, Christopher E.; Noble, Peter A.; Milliken, Charles E.; Li, Dien; Kaplan, Daniel I.

2018-01-01

Water quality is an important determinant for the structural integrity of alloy cladded fuels and assemblies during long-term wet storage. Detailed characterization of a water filled storage basin for spent nuclear reactor fuel was performed following the formation and proliferation of an amorphous white flocculent. White precipitant was sampled throughout the storage basin for chemical and spectroscopic characterization, and environmental DNA was extracted for 454 pyrosequencing of bacterial 16S rRNA gene diversity. Accordingly, spectroscopic analyses indicated the precipitant to be primarily amorphous to crystalline aluminum (oxy) hydroxides with minor associated elemental components including Fe, Si, Ti, and U. High levels of organic carbon were co-localized with the precipitant relative to bulk dissolved organic concentrations. Bacterial densities were highly variable between sampling locations and with depth within the water filled storage basin; cell numbers ranged from 4 × 103to 4 × 104 cells/mL. Bacterial diversity that was physically associated with the aluminum (oxy) hydroxide complexes exceeded an estimated 4,000 OTUs/amplicon library (3% cutoff) and the majority of sequences were aligned to the families Burkholderiaceae (23%), Nitrospiraceae (23%), Hyphomicrobiaceae (17%), and Comamonadaceae (6%). We surmise that episodic changes in the physical and chemical properties of the basin contribute to the polymerization of aluminum (oxy) hydroxides, which in turn can chemisorb nutrients, carbon ligands and bacterial cells from the surrounding bulk aqueous phase. As such, these precipitants should establish favorable microhabitats for bacterial colonization and growth. Comparative analyses of 16S rRNA gene amplicon libraries across a selection of natural and engineered aquatic ecosystems were performed and microbial community and taxonomic signatures unique to the spent nuclear fuel (SNF) storage basin environment were revealed. These insights could spur the

FY13 Summary Report on the Augmentation of the Spent Fuel Composition Dataset for Nuclear Forensics: SFCOMPO/NF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brady Raap, Michaele C.; Lyons, Jennifer A.; Collins, Brian A.

This report documents the FY13 efforts to enhance a dataset of spent nuclear fuel isotopic composition data for use in developing intrinsic signatures for nuclear forensics. A review and collection of data from the open literature was performed in FY10. In FY11, the Spent Fuel COMPOsition (SFCOMPO) excel-based dataset for nuclear forensics (NF), SFCOMPO/NF was established and measured data for graphite production reactors, Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) were added to the dataset and expanded to include a consistent set of data simulated by calculations. A test was performed to determine whether the SFCOMPO/NF dataset willmore » be useful for the analysis and identification of reactor types from isotopic ratios observed in interdicted samples.« less

Spent Nuclear Fuel Disposition

DOE PAGES

Wagner, John C.

2016-05-22

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

Spent Nuclear Fuel Disposition

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagner, John C.

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paskevych, Sergiy; Batiy, Valiriy; Sizov, Andriy

2013-07-01

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

Thermodynamics of Electron Flow in the Bacterial Deca-heme Cytochrome MtrF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Breuer, Marian; Zarzycki, Piotr P.; Blumberger, Jochen

2012-07-01

Electron transporting multiheme cytochromes are essential to the metabolism of microbes that inhabit soils and carry out important biogeochemical processes. Recently the first crystal structure of a prototype bacterial deca-heme cytochrome (MtrF) has been resolved and its electrochemistry characterized. However, the molecular details of electron conductance along heme chains in the cytochrome are difficult to access via experiment due to the nearly identical chemical nature of the heme cofactors. Here we employ large-scale molecular dynamics simulations to compute the reduction potentials of the ten hemes of MtrF in aqueous solution. We find that as a whole they fall within amore » range of about 0.3 V in agreement with experiment. Individual reduction potentials give rise to a free energy profile for electron conduction that is approximately symmetric with respect to the center of the protein. Our calculations indicate that there is no significant potential bias along the orthogonal octa- and tetra-heme chains suggesting that under aqueous conditions MtrF is a nearly reversible two-dimensional conductor.« less

Flavin binding to the deca-heme cytochrome MtrC: Insights from computational molecular simulation

DOE PAGES

Breuer, Marian; Rosso, Kevin  M.; Blumberger, Jochen

2015-12-15

Here, certain dissimilatory bacteria have the remarkable ability to use extracellular metal oxide minerals instead of oxygen as terminal electron sinks, using a process known as “extracellular respiration”. Specialized multiheme cytochromes located on the outer membrane of the microbe were shown to be crucial for electron transfer from the cell surface to the mineral. This process is facilitated by soluble, biogenic flavins secreted by the organism for the purpose of acting as an electron shuttle. However, their interactions with the outer-membrane cytochromes are not established on a molecular scale. Here, we study the interaction between the outer-membrane deca-heme cytochrome MtrCmore » from Shewanella oneidensis and flavin mononucleotide (FMN in fully oxidized quinone form) using computational docking. We find that interaction of FMN with MtrC is significantly weaker than with known FMN-binding proteins, but identify a mildly preferred interaction site close to heme 2 with a dissociation constant (K d) = 490 μM, in good agreement with recent experimental estimates, K d = 255 μM. The weak interaction with MtrC can be qualitatively explained by the smaller number of hydrogen bonds that the planar headgroup of FMN can form with this protein compared to FMN-binding proteins. Molecular dynamics simulation gives indications for a possible conformational switch upon cleavage of the disulphide bond of MtrC, but without concomitant increase in binding affinities according to this docking study. Overall, our results suggest that binding of FMN to MtrC is reversible and not highly specific, which may be consistent with a role as redox shuttle that facilitates extracellular respiration.« less

MTR,TRA603. EXPERIMENTERS' SPACE ALLOCATIONS IN BASEMENT AS OF 1963. SHIELDED ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR,TRA-603. EXPERIMENTERS' SPACE ALLOCATIONS IN BASEMENT AS OF 1963. SHIELDED CUBICLES WERE IDENTIFIED BY SPONSORING LABORATORY AND ITS TEST HOLE NUMBER IN THE REACTOR, IE, "KAPL HB-1" SIGNIFIED KNOLLS ATOMIC POWER LABORATORY, HORIZONTAL BEAM NO. 1. "WAPD" WAS WESTINGHOUSE ATOMIC POWER DIVISION. CATCH TANKS AND SAMPLE STATIONS FOR TEST LOOPS WERE ASSOCIATED WITH THESE CUBICLES. NOTE DESKS, STORAGE CABINETS, SWITCH GEAR, INSTRUMENT PANELS. PHILLIPS PETROLEUM COMPANY MTR-E-5205, 4/1963. INL INDEX NO. 531-0603-00-706-009757, REV. 5. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Incorporation mechanisms of actinide elements into the structures of U 6+ phases formed during the oxidation of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Burns, Peter C.; Ewing, Rodney C.; Miller, Mark L.

1997-05-01

Uranyl oxide hydrate and uranyl silicate phases will form due to the corrosion and alteration of spent nuclear fuel under oxidizing conditions in silica-bearing solution. The actinide elements in the spent fuel may be incorporated into the structures of these secondary U6+ phases during the long-term corrosion of the UO 2 in spent fuel. The incorporation of actinide elements into the crystal structures of the alteration products may decrease actinide mobility. The crystal chemistry of the various oxidation states of the actinide elements of environmental concern is examined to identify possible incorporation mechanisms. The substitutions Pu 6+U 6+ and (Pu 5+, Np 5+)U 6+ should readily occur in many U 6+ structures, although structural modification may be required to satisfy local bond-valence requirements. Crystal-chemical characteristics of the U 6+ phases indicate that An 4+ (An: actinide)U 6+ substitution is likely to occur in the sheets of uranyl polyhedra that occur in the structures of the minerals schoepite, [(UO 2) 8O 2(OH) 12](H 2O) 12, ianthinite, [U 24+ (UO 2) 4O 6(OH) 4(H 2O) 4](H 2O) 5, becquerelite, Ca[(UO 2) 3O 2(OH) 3] 2(H 2O) 8, compreignacite, K 2[(UO 2) 3O 2(OH) 3] 2(H 2O) 8, α-uranophane, Ca[(UO 2)(SiO 3OH)] 2(H 2O) 5, and boltwoodite, K(H 2O)[(UO 2)(SiO 4)], all of which are likely to form due to the oxidation and alteration of the UO 2 in spent fuel. The incorporation of An 3+ into the sheets of the structures of α-uranophane and boltwoodite, as well as interlayer sites of various uranyl phases, may occur.

Sequestration of radioactive iodine in silver-palladium phases in commercial spent nuclear fuel

NASA Astrophysics Data System (ADS)

Buck, Edgar C.; Mausolf, Edward J.; McNamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

2016-12-01

Radioactive iodine is the Achilles' heel in the design for the safe geological disposal of spent uranium oxide (UO2) nuclear fuel. Furthermore, iodine's high volatility and aqueous solubility were mainly responsible for the high early doses released during the accident at Fukushima Daiichi in 2011. Studies Kienzler et al., however, have indicated that the instant release fraction (IRF) of radioiodine (131/129I) does not correlate directly with increasing fuel burn-up. In fact, there is a peak in the release of iodine at around 50-60 MW d/kgU, and with increasing burn-up, the IRF of 131/129I decreases. The reasons for this decrease have not fully been understood. We have performed microscopic analysis of chemically processed high burn-up UO2 fuel (80 MW d/kgU) and have found recalcitrant nano-particles containing, Pd, Ag, I, and Br, possibly consistent with a high pressure phase of silver iodide in the undissolved residue. It is likely that increased levels of Ag and Pd from 239Pu fission in high burnup fuels leads to the formation of these metal halides. The occurrence of these phases in UO2 nuclear fuels may reduce the impact of long-lived 129I on the repository performance assessment calculations.

PRECAST CONCRETE WALL PANELS ARE LIFTED INTO PLACE ON MTR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PRECAST CONCRETE WALL PANELS ARE LIFTED INTO PLACE ON MTR STEEL FRAME STRUCTURE. INL NEGATIVE NO. 1330. Unknown Photographer, 1/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

NDE of copper canisters for long-term storage of spent nuclear fuel from the Swedish nuclear power plants

NASA Astrophysics Data System (ADS)

Stepinski, Tadeusz

2003-07-01

Sweden has been intensively developing methods for long term storage of spent fuel from the nuclear power plants for twenty-five years. A dedicated research program has been initiated and conducted by the Swedish company SKB (Swedish Nuclear Fuels and Waste Management Co.). After the interim storage SKB plans to encapsulate spent nuclear fuel in copper canisters that will be placed at a deep repository located in bedrock. The canisters filled with fuel rods will be sealed by an electron beam weld. This paper presents three complementary NDE techniques used for assessing the sealing weld in copper canisters, radiography, ultrasound, and eddy current. A powerful X-ray source and a digital detector are used for the radiography. An ultrasonic array system consisting of a phased ultrasonic array and a multi-channel electronics is used for the ultrasonic examination. The array system enables electronic focusing and rapid electronic scanning eliminating the use of a complicated mechanical scanner. A specially designed eddy current probe capable of detecting small voids at the depth up to 4 mm in copper is used for the eddy current inspection. Presently, all the NDE techniques are verified in SKB's Canister Laboratory where full scale canisters are welded and examined.

The preliminary design and feasibility study of the spent fuel and high level waste repository in the Czech Republic

DOE Office of Scientific and Technical Information (OSTI.GOV)

Valvoda, Z.; Holub, J.; Kucerka, M.

1996-12-31

In the year 1993, began the Program of Development of the Spent Fuel and High Level Waste Repository in the Conditions of the Czech Republic. During the first phase, the basic concept and structure of the Program has been developed, and the basic design criteria and requirements were prepared. In the conditions of the Czech Republic, only an underground repository in deep geological formation is acceptable. Expected depth is between 500 to 1000 meters and as host rock will be granites. A preliminary variant design study was realized in 1994, that analyzed the radioactive waste and spent fuel flow frommore » NPPs to the repository, various possibilities of transportation in accordance to the various concepts of spent fuel conditioning and transportation to the underground structures. Conditioning and encapsulation of spent fuel and/or radioactive waste is proposed on the repository site. Underground disposal structures are proposed at one underground floor. The repository will have reserve capacity for radioactive waste from NPPs decommissioning and for waste non acceptable to other repositories. Vertical disposal of unshielded canisters in boreholes and/or horizontal disposal of shielded canisters is studied. As the base term of the start up of the repository operation, the year 2035 has been established. From this date, a preliminary time schedule of the Project has been developed. A method of calculating leveled and discounted costs within the repository lifetime, for each of selected 5 variants, was used for economic calculations. Preliminary expected parametric costs of the repository are about 0,1 Kc ($0.004) per MWh, produced in the Czech NPPs. In 1995, the design and feasibility study has gone in more details to the technical concept of repository construction and proposed technologies, as well as to the operational phase of the repository. Paper will describe results of the 1995 design work and will present the program of the repository development in

MTR BASEMENT. GENERAL ELECTRIC CONTROL CONSOLE FOR AIRCRAFT NUCLEAR PROPULSION ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR BASEMENT. GENERAL ELECTRIC CONTROL CONSOLE FOR AIRCRAFT NUCLEAR PROPULSION EXPERIMENT NO. 1. INL NEGATIVE NO. 6510. Unknown Photographer, 9/29/1959 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-07-01

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

Neutronic study on conversion of SAFARI-1 to LEU silicide fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ball, G.; Pond, R.; Hanan, N.

1995-02-01

This paper marks the initial study into the technical and economic feasibility of converting the SAFARI-1 reactor in South Africa to LEU silicide fuel. Several MTR assembly geometries and LEU uranium densities have been studied and compared with MEU and HEU fuels. Two factors of primary importance for conversion of SAFARI-1 to LEU fuel are the economy of the fuel cycle and the performance of the incore and excore irradiation positions.

Roles of Radiolytic and Externally Generated H2 in the Corrosion of Fractured Spent Nuclear Fuel.

PubMed

Liu, Nazhen; Wu, Linda; Qin, Zack; Shoesmith, David W

2016-11-15

A 2-D model for the corrosion of spent nuclear fuel inside a failed nuclear waste container has been modified to determine the influence of various redox processes occurring within fractures in the fuel. The corrosion process is driven by reaction of the fuel with the dominant α radiolysis product, H 2 O 2 . A number of reactions are shown to moderate or suppress the corrosion rate, including H 2 O 2 decomposition and a number of reactions involving dissolved H 2 produced either by α radiolysis or by the corrosion of the steel container vessel. Both sources of H 2 lead to the suppression of fuel corrosion, with their relative importance being determined by the radiation dose rate, the steel corrosion rate, and the dimensions of the fractures in the fuel. The combination of H 2 from these two sources can effectively prevent corrosion when only micromolar quantities of H 2 are present.

MTR CONTROL ROOM WITH CONTROL CONSOLE AND STATUS READOUTS ALONG ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR CONTROL ROOM WITH CONTROL CONSOLE AND STATUS READOUTS ALONG WALL. WORKERS MAKE ELECTRICAL AND OTHER CONNECTIONS. INL NEGATIVE NO. 4289. Unknown Photographer, 2/26/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Utilization of the Differential Die-Away Self-Interrogation Technique for Characterization and Verification of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Trahan, Alexis Chanel

New nondestructive assay techniques are sought to better characterize spent nuclear fuel. One of the NDA instruments selected for possible deployment is differential die-away self-interrogation (DDSI). The proposed DDSI approach for spent fuel assembly assay utilizes primarily the spontaneous fission and (alpha, n) neutrons in the assemblies as an internal interrogating radiation source. The neutrons released in spontaneous fission or (alpha,n) reactions are thermalized in the surrounding water and induce fission in fissile isotopes, thereby creating a measurable signal from isotopes of interest that would be otherwise difficult to measure. The DDSI instrument employs neutron coincidence counting with 3He tubes and list-mode-based data acquisition to allow for production of Rossi-alpha distributions (RADs) in post-processing. The list-mode approach to data collection and subsequent construction of RADs has expanded the analytical possibilities, as will be demonstrated throughout this thesis. One of the primary advantages is that the measured signal in the form of a RAD can be analyzed in its entirety including determination of die-away times in different time domains. This capability led to the development of the early die-away method, a novel leakage multiplication determination method which is tested throughout the thesis on different sources in simulation space and fresh fuel experiments. The early die-away method is a robust, accurate, improved method of determining multiplication without the need for knowledge of the (alpha,n) source term. The DDSI technique and instrument are presented along with the many novel capabilities enabled by and discovered through RAD analysis. Among the new capabilities presented are the early die-away method, total plutonium content determination, and highly sensitive missing pin detection. Simulation of hundreds of different spent and fresh fuel assemblies were used to develop the analysis algorithms and the techniques were

Utilization of the Differential Die-Away Self-Interrogation Technique for Characterization and Verification of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trahan, Alexis Chanel

New nondestructive assay techniques are sought to better characterize spent nuclear fuel. One of the NDA instruments selected for possible deployment is differential die-away self-interrogation (DDSI). The proposed DDSI approach for spent fuel assembly assay utilizes primarily the spontaneous fission and (α, n) neutrons in the assemblies as an internal interrogating radiation source. The neutrons released in spontaneous fission or (α,n) reactions are thermalized in the surrounding water and induce fission in fissile isotopes, thereby creating a measurable signal from isotopes of interest that would be otherwise difficult to measure. The DDSI instrument employs neutron coincidence counting with 3He tubesmore » and list-mode-based data acquisition to allow for production of Rossi-alpha distributions (RADs) in post-processing. The list-mode approach to data collection and subsequent construction of RADs has expanded the analytical possibilities, as will be demonstrated throughout this thesis. One of the primary advantages is that the measured signal in the form of a RAD can be analyzed in its entirety including determination of die-away times in different time domains. This capability led to the development of the early die-away method, a novel leakage multiplication determination method which is tested throughout the thesis on different sources in simulation space and fresh fuel experiments. The early die-away method is a robust, accurate, improved method of determining multiplication without the need for knowledge of the (α,n) source term. The DDSI technique and instrument are presented along with the many novel capabilities enabled by and discovered through RAD analysis. Among the new capabilities presented are the early die-away method, total plutonium content determination, and highly sensitive missing pin detection. Simulation of hundreds of different spent and fresh fuel assemblies were used to develop the analysis algorithms and the techniques were tested

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE PAGES

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.; ...

2017-08-02

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

YUCCA Mountain Project - Argonne National Laboratory, Annual Progress Report, FY 1997 for activity WP 1221 unsaturated drip condition testing of spent fuel and unsaturated dissolution tests of glass.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bates, J. K.; Buck, E. C.; Emery, J. W.

1998-09-18

This document reports on the work done by the Nuclear Waste Management Section of the Chemical Technology Division of Argonne National Laboratory in the period of October 1996 through September 1997. Studies have been performed to evaluate the behavior of nuclear waste glass and spent fuel samples under the unsaturated conditions (low-volume water contact) that are likely to exist in the Yucca Mountain environment being considered as a potential site for a high-level waste repository. Tests with actinide-doped waste glasses, in progress for over 11 years, indicate that the transuranic element release is dominated by colloids that continuously form andmore » span from the glass surface. The nature of the colloids that form in the glass and spent fuel testing programs is being investigated by dynamic light scattering to determine the size distribution, by autoradiography to determine the chemistry, and by zeta potential to measure the electrical properties of the colloids. Tests with UO{sub 2} have been ongoing for 12 years. They show that the oxidation of UO{sub 2} occurs rapidly, and the resulting paragenetic sequence of secondary phases forming on the sample surface is similar to that observed for uranium found in natural oxidizing environments. The reaction of spent fuel samples in conditions similar to those used with UO{sub 2} have been in progress for over six years, and the results suggest that spent fuel forms many of the same alteration products as UO{sub 2}. With spent fuel, the bulk of the reaction occurs via a through-grain reaction process, although grain boundary attack is sufficient to have reacted all of the grain boundary regions in the samples. New test methods are under development to evaluate the behavior of spent fuel samples with intact cladding: the rate at which alteration and radionuclide release occurs when water penetrates fuel sections and whether the reaction causes the cladding to split. Alteration phases have been formed on fine

10 CFR 51.23 - Temporary storage of spent fuel after cessation of reactor operation-generic determination of no...

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Temporary storage of spent fuel after cessation of reactor operation-generic determination of no significant environmental impact. 51.23 Section 51.23 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC LICENSING AND RELATED...

Water Chemistry Control System for Recovery of Damaged and Degraded Spent Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sindelar, R.; Fisher, D.; Thomas, J.

2011-02-18

The International Atomic Energy Agency (IAEA) and the government of Serbia have led the project cosponsored by the U.S, Russia, European Commission, and others to repackage and repatriate approximately 8000 spent fuel elements from the RA reactor fuel storage basins at the VIN?A Institute of Nuclear Sciences to Russia for reprocessing. The repackaging and transportation activities were implemented by a Russian consortium which includes the Sosny Company, Tekhsnabeksport (TENEX) and Mayak Production Association. High activity of the water of the fuel storage basin posed serious risk and challenges to the fuel removal from storage containers and repackaging for transportation. Themore » risk centered on personnel exposure, even above the basin water, due to the high water activity levels caused by Cs-137 leached from fuel elements with failed cladding. A team of engineers from the U.S. DOE-NNSA's Global Threat Reduction Initiative, the Vinca Institute, and the IAEA performed the design, development, and deployment of a compact underwater water chemistry control system (WCCS) to remove the Cs-137 from the basin water and enable personnel safety above the basin water for repackaging operations. Key elements of the WCCS system included filters, multiple columns containing an inorganic sorbent, submersible pumps and flow meters. All system components were designed to be remotely serviceable and replaceable. The system was assembled and successfully deployed at the Vinca basin to support the fuel removal and repackaging activities. Following the successful operations, the Cs-137 is now safely contained and consolidated on the zeolite sorbent used in the columns of the WCCS, and the fuel has been removed from the basins. This paper reviews the functional requirements, design, and deployment of the WCCS.« less

Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel and Transuranic Radioactive Wastes (40 CFR Part 191)

EPA Pesticide Factsheets

This regulation sets environmental standards for public protection from the management and disposal of spent nuclear fuel, high-level wastes and wastes that contain elements with atomic numbers higher than uranium (transuranic wastes).

Recovery of minor actinides from spent fuel using TPEN-immobilized gels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koyama, S.; Suto, M.; Ohbayashi, H.

2013-07-01

A series of separation experiments was performed in order to study the recovery process for minor actinides (MAs), such as americium (Am) and curium (Cm), from the actual spent fuel by using an extraction chromatographic technique. N,N,N',N'-tetrakis-(4-propenyloxy-2-pyridylmethyl) ethylenediamine (TPPEN) is an N,N,N',N'-tetrakis (2-pyridylmethyl) ethylenediamine (TPEN) analogue consisting of an incorporated pyridine ring that acts as not only a ligand but also as a site for polymerization and crosslinking of the gel. The TPPEN and N-isopropylacrylamide (NIPA) were dissolved into dimethylformamide (DMF, Wako Co., Ltd.) and a silica beads polymer, and then TTPEN was immobilized chemically in a polymer gel (somore » called TPEN-gel). Mixed oxide (MOX) fuel, which was highly irradiated up to 119 GWD/MTM in the experimental fast reactor Joyo, was used as a reference spent fuel. First, uranium (U) and plutonium (Pu) were separated from the irradiated fuel using an ion-exchange method, and then, the platinum group elements were removed by CMPO to leave a mixed solution of MAs and lanthanides. The 3 mol% TPPEN-gel was packed with as an extraction column (CV: 1 ml) and then rinsed by 0.1 M NaNO{sub 3}(pH 4.0) for pH adjustment. After washing the column by 0.01 M NaNO{sub 3} (pH 4.0), Eu was detected and the recovery rate reached 93%. The MAs were then recovered by changing the eluent to 0.01 M NaNO{sub 3} (pH 2.0), and the recovery rate of Am was 48 %. The 10 mol% TPPEN-gel was used to improve adsorption coefficient of Am and a condition of eluent temperature was changed in order to confirm the temperature swing effect on TPEN-gel for MA. More than 90% Eu was detected in the eluent after washing with 0.01 M NaNO{sub 3} (pH 3.5) at 5 Celsius degrees. Americium was backwardly detected and eluted continuously during the same condition. After removal of Eu, the eluent temperature was changed to 32 Celsius degrees, then Am was detected (pH 3.0). Finally remained Am could be

Extended Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ade, Brian J; Bowman, Stephen M; Gauld, Ian C

2015-01-01

[Full Text] Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (k eff) calculations and depleted fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date, investigating some aspects of extended BUC, andmore » it also describes the plan to complete the evaluations. The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper. Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC, including investigation of the axial void profile effect and the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of an operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. While a single cycle does not provide complete data, the data obtained are sufficient to use to determine the primary effects and identify conservative modeling approaches. Using data resulting from a single cycle, the axial void profile is studied by first determining the temporal fidelity necessary in depletion modeling, and then using multiple void profiles to examine the effect of the void profile on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied is control blade exposure. Control blades

Impact of Americium-241 (n,γ) Branching Ratio on SFR Core Reactivity and Spent Fuel Characteristics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hiruta, Hikaru; Youinou, Gilles J.; Dixon, Brent W.

An accurate prediction of core physics and fuel cycle parameters largely depends on the order of details and accuracy in nuclear data taken into account for actual calculations. 241Am is a major gateway nuclide for most of minor actinides and thus important nuclide for core physics and fuel-cycle calculations. The 241Am(n,?) branching ratio (BR) is in fact the energy dependent (see Fig. 1), therefore, it is necessary to taken into account the spectrum effect on the calculation of the average BR for the full-core depletion calculations. Moreover, the accuracy of the BR used in the depletion calculations could significantly influencemore » the core physics performance and post irradiated fuel compositions. The BR of 241Am(n,?) in ENDF/B-VII.0 library is relatively small and flat in thermal energy range, gradually increases within the intermediate energy range, and even becomes larger at the fast energy range. This indicates that the properly collapsed BR for fast reactors could be significantly different from that of thermal reactors. The evaluated BRs are also differ from one evaluation to another. As seen in Table I, average BRs for several evaluated libraries calculated by means of a fast spectrum are similar but have some differences. Most of currently available depletion codes use a pre-determined single value BR for each library. However, ideally it should be determined on-the-fly basis like that of one-group cross sections. These issues provide a strong incentive to investigate the effect of different 241Am(n,?) BRs on core and spent fuel parameters. This paper investigates the impact of the 241Am(n,?) BR on the results of SFR full-core based fuel-cycle calculations. The analysis is performed by gradually increasing the value of BR from 0.15 to 0.25 and studying its impact on the core reactivity and characteristics of SFR spent fuels over extended storage times (~10,000 years).« less

Structure and reconstitution of yeast Mpp6-nuclear exosome complexes reveals that Mpp6 stimulates RNA decay and recruits the Mtr4 helicase.

PubMed

Wasmuth, Elizabeth V; Zinder, John C; Zattas, Dimitrios; Das, Mom; Lima, Christopher D

2017-07-25

Nuclear RNA exosomes catalyze a range of RNA processing and decay activities that are coordinated in part by cofactors, including Mpp6, Rrp47, and the Mtr4 RNA helicase. Mpp6 interacts with the nine-subunit exosome core, while Rrp47 stabilizes the exoribonuclease Rrp6 and recruits Mtr4, but it is less clear if these cofactors work together. Using biochemistry with Saccharomyces cerevisiae proteins, we show that Rrp47 and Mpp6 stimulate exosome-mediated RNA decay, albeit with unique dependencies on elements within the nuclear exosome. Mpp6-exosomes can recruit Mtr4, while Mpp6 and Rrp47 each contribute to Mtr4-dependent RNA decay, with maximal Mtr4-dependent decay observed with both cofactors. The 3.3 Å structure of a twelve-subunit nuclear Mpp6 exosome bound to RNA shows the central region of Mpp6 bound to the exosome core, positioning its Mtr4 recruitment domain next to Rrp6 and the exosome central channel. Genetic analysis reveals interactions that are largely consistent with our model.

Structure and reconstitution of yeast Mpp6-nuclear exosome complexes reveals that Mpp6 stimulates RNA decay and recruits the Mtr4 helicase

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wasmuth, Elizabeth V.; Zinder, John C.; Zattas, Dimitrios

Nuclear RNA exosomes catalyze a range of RNA processing and decay activities that are coordinated in part by cofactors, including Mpp6, Rrp47, and the Mtr4 RNA helicase. Mpp6 interacts with the nine-subunit exosome core, while Rrp47 stabilizes the exoribonuclease Rrp6 and recruits Mtr4, but it is less clear if these cofactors work together. Using biochemistry with Saccharomyces cerevisiae proteins, we show that Rrp47 and Mpp6 stimulate exosome-mediated RNA decay, albeit with unique dependencies on elements within the nuclear exosome. Mpp6-exosomes can recruit Mtr4, while Mpp6 and Rrp47 each contribute to Mtr4-dependent RNA decay, with maximal Mtr4-dependent decay observed with bothmore » cofactors. The 3.3 Å structure of a twelve-subunit nuclear Mpp6 exosome bound to RNA shows the central region of Mpp6 bound to the exosome core, positioning its Mtr4 recruitment domain next to Rrp6 and the exosome central channel. Genetic analysis reveals interactions that are largely consistent with our model.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1996-02-01

On June 1, 1995, DOE issued a Record of Decision [60 Federal Register 28680] for the Department-wide management of spent nuclear fuel (SNF); regionalized storage of SNF by fuel type was selected as the preferred alternative. The proposed action evaluated in this environmental assessment is the management of SNF on the Oak Ridge Reservation (ORR) to implement this preferred alternative of regional storage. SNF would be retrieved from storage, transferred to a hot cell if segregation by fuel type and/or repackaging is required, loaded into casks, and shipped to off-site storage. The proposed action would also include construction and operationmore » 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.« less

Comparative analyses of spent nuclear fuel transport modal options: Transport options under existing site constraints

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brentlinger, L.A.; Hofmann, P.L.; Peterson, R.W.

1989-08-01

The movement of nuclear waste can be accomplished by various transport modal options involving different types of vehicles, transport casks, transport routes, and intermediate intermodal transfer facilities. A series of systems studies are required to evaluate modal/intermodal spent fuel transportation options in a consistent fashion. This report provides total life-cycle cost and life-cycle dose estimates for a series of transport modal options under existing site constraints. 14 refs., 7 figs., 28 tabs.

Targeted Protein Degradation of Outer Membrane Decaheme Cytochrome MtrC Metal Reductase in Shewanella oneidensis MR-1 Measured Using Biarsenical Probe CrAsH-EDT2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xiong, Yijia; Chen, Baowei; Shi, Liang

2011-10-14

Development of efficient microbial biofuel cells requires an ability to exploit interfacial electron transfer reactions to external electron acceptors, such as metal oxides; such reactions occur in the facultative anaerobic gram-negative bacterium Shewanella oneidensis MR-1 through the catalytic activity of the outer membrane decaheme c-type cytochrome MtrC. Central to the utility of this pathway to synthetic biology is an understanding of cellular mechanisms that maintain optimal MtrC function, cellular localization, and renewal by degradation and resynthesis. In order to monitor trafficking to the outer membrane, and the environmental sensitivity of MtrC, we have engineered a tetracysteine tag (i.e., CCPGCC) atmore » its C-terminus that permits labeling by the cell impermeable biarsenical fluorophore, carboxy-FlAsH (CrAsH) of MtrC at the surface of living Shewanella oneidensis MR-1 cells. In comparison, the cell permeable reagent FlAsH permits labeling of the entire population of MtrC, including proteolytic fragments resulting from incorrect maturation. We demonstrate specific labeling by CrAsH of engineered MtrC which is dependent on the presence of a functional type-2 secretion system (T2S), as evidenced by T2S system gspD or gspG deletion mutants which are incapable of CrAsH labeling. Under these latter conditions, MtrC undergoes proteolytic degradation to form a large 35-38 kDa fragment; this degradation product is also resolved during normal turnover of the CrAsH-labeled MtrC protein. No MtrC protein is released into the medium during turnover, suggesting the presence of cellular turnover systems involving MtrC reuptake and degradation. The mature MtrC localized on the outer membrane is a long-lived protein, with a turnover rate of 0.043 hr-1 that is insensitive to O2 concentration. Maturation of MtrC is relatively inefficient, with substantial rates of turnover of the immature protein prior to export to the outer membrane (i.e., 0.028 hr-1) that are

Accelerator-Driven Subcritical System for Disposing of the U.S. Spent Nuclear Fuel Inventory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gohar, Yousry; Cao, Yan; Kraus, Adam R.

The current United States inventory of the spent nuclear fuel (SNF) is ~80,000 metric tons of heavy metal (MTHM), including ~131 tons of minor actinides (MAs) and ~669 tons of plutonium. This study describes a conceptual design of an accelerator-driven subcritical (ADS) system for disposing of this SNF inventory by utilizing the 131 tons of MAs inventory and a fraction of the plutonium inventory for energy production, and transmuting some long-lived fission products. An ADS system with a homogeneous subcritical fission blanket was first examined. A spallation neutron source is used to drive the blanket and it is produced frommore » the interaction of a 1-GeV proton beam with a lead-bismuth eutectic (LBE) target. The blanket has a liquid mobile fuel using LBE as the fuel carrier. The fuel materials are dissolved, mixed, or suspended in the liquid fuel carrier. Monte Carlo analyses were performed to determine the overall parameters of the concept. Steady-state Monte Carlo simulations were performed for three similar fission blankets. Except for, the loaded amount of actinide materials in the LBE is either 5, 7, or 10% of the total volume of the blanket, respectively. The neutron multiplication factors of the three blankets are ~0.98 and the initial MAs blanket inventories are ~10 tons. In addition, Monte Carlo burnup simulations using the MCB5 code were performed to analyze the performance of the three conceptual ADS systems. During operation, fresh fuel was fed into the fission blanket to adjust its reactivity and to control the system power. The burnup analysis shows that the three ADS concepts consume about 1.2 tons of actinides per full power year and produce 3 GW thermal power, with a proton beam power of 25 MW. For the blankets with 5, 7, or 10% actinide fuel particles loaded in the LBE, assuming that the ADS systems can be operated for 35 full-power years, the total MA materials consumed in the three ADS systems are about 30.6, 35.3, and 37.2 tons

Planning, Preparation, and Transport of the High-Enriched Uranium Spent Nuclear Fuel from the Czech Republic to the Russian Federation

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. J. Tyacke; I. Bolshinsky; Frantisek Svitak

The United States, Russian Federation, and the International Atomic Energy Agency have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program, which is part of the Global Threat Reduction Initiative. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. In February 2003, the RRRFR Program began discussions with the Nuclear Research Institute (NRI) in Rež, Czech Republic, about returning their HEU spent nuclear fuel to the Russian Federation for reprocessing. In March 2005, the U.S. Department ofmore » Energy signed a contract with NRI to perform all activities needed for transporting their HEU spent nuclear fuel to Russia. After 2 years of intense planning, preparations, and coordination at NRI and with three other countries, numerous organizations and agencies, and a Russian facility, this shipment is scheduled for completion before the end of 2007. This paper will provide a summary of activities completed for making this international shipment. This paper contains an introduction and background of the RRRFR Program and the NRI shipment project. It summarizes activities completed in preparation for the shipment, including facility preparations at NRI in Rež and FSUE “Mayak” in Ozyorsk, Russia; a new transportation cask system; regulatory approvals; transportation planning and preparation in the Czech Republic, Slovakia, Ukraine, and the Russian Federation though completion of the Unified Project and Special Ecological Programs. The paper also describes fuel loading and cask preparations at NRI and final preparations/approvals for transporting the shipment across the Czech Republic, Slovakia, Ukraine, and the Russian Federation to FSUE Mayak where the HEU spent nuclear fuel will be processed, the uranium will be downblended and made into low-enriched uranium fuel for commercial

SFCOMPO 2.0 - A relational database of spent fuel isotopic measurements, reactor operational histories, and design data

NASA Astrophysics Data System (ADS)

Michel-Sendis, Franco; Martinez-González, Jesus; Gauld, Ian

2017-09-01

SFCOMPO-2.0 is a database of experimental isotopic concentrations measured in destructive radiochemical analysis of spent nuclear fuel (SNF) samples. The database includes corresponding design description of the fuel rods and assemblies, relevant operating conditions and characteristics of the host reactors necessary for modelling and simulation. Aimed at establishing a thorough, reliable, and publicly available resource for code and data validation of safety-related applications, SFCOMPO-2.0 is developed and maintained by the OECD Nuclear Energy Agency (NEA). The SFCOMPO-2.0 database is a Java application which is downloadable from the NEA website.

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR). Data will be collected under simulated transportation environments using the cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). These data will be used to support ongoing SNF modeling activities and to address regulatory issues associated with SNF transport.

Valorization of spent coffee grounds recycling as a potential alternative fuel resource in Turkey: An experimental study.

PubMed

Atabani, A E; Mercimek, S M; Arvindnarayan, Sundaram; Shobana, Sutha; Kumar, Gopalakrishnan; Cadir, Mehmet; Al-Muhatseb, Ala'a H

2018-03-01

In this study, recycling of spent coffee grounds (SCG) as a potential feedstock for alternative fuel production and compounds of added value in Turkey was assessed. The average oil content was found (≈ 13% w/w). All samples (before and after extraction) were tested for scanning electron microscopy (SEM), differential scanning calorimetry (DSC), thermogravimetric analysis (TGA), X-ray diffraction (XRD), calorific value, surface analysis and porosity, Fourier transform infrared (FT-IR), and elemental analysis to assess their potential towards fuel properties. Elemental analysis indicated that carbon represents the highest percentages (49.59% and 46.42%, respectively), followed by nitrogen (16.7% and 15.5%), hydrogen (6.74% and 6.04%), and sulfur (0.851% and 0.561%). These results indicate that SCG can be utilized as compost, as it is rich in nitrogen. Properties of the extracted oil were examined, followed by biodiesel production. The quality of biodiesel was compared with American Society for Testing and Materials (ASTM) D6751 standards, and all the properties complied with standard specifications. The fatty acid compositions were analyzed by gas chromatography. It was observed that coffee waste methyl ester (CWME) is mainly composed of palmitic (35.8%) and arachidic (44.6%) acids, which are saturated fatty acids. The low degree of unsaturation provides an excellent oxidation stability (10.4 hr). CWME has also excellent cetane number, higher heating value, and iodine value with poor cold flow properties. The studies also investigated blending of biodiesel with Euro diesel and butanol. Following this, a remarkable improvement in cloud and pour points of biodiesel was obtained. Spent coffee grounds after oil extraction is an ideal material for garden fertilizer, feedstock for ethanol, biogas production, and as fuel pellets. The outcome of such research work produces valuable insights on the recycling importance of SCG in Turkey. Coffee is a huge industry, and

SOUTH WING, TRA661. SOUTH SIDE. CAMERA FACING NORTH. MTR HIGH ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

SOUTH WING, TRA-661. SOUTH SIDE. CAMERA FACING NORTH. MTR HIGH BAY BEYOND. INL NEGATIVE NO. HD46-45-3. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

75 FR 57841 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6, Confirmation of Effective...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-23

... Spent Fuel Storage Casks: NAC-MPC System, Revision 6, Confirmation of Effective Date AGENCY: Nuclear... include Amendment Number 6 to Certificate of Compliance (CoC) Number 1025. DATES: Effective Date: The... regulations at 10 CFR 72.214 to include Amendment No. 6 to CoC No. 1025. Amendment No. 6 changes the...

A discussion on the methodology for calculating radiological and toxicological consequences for the spent nuclear fuel project at the Hanford Site

DOE Office of Scientific and Technical Information (OSTI.GOV)

RITTMANN, P.D.

1999-07-14

This report contains technical information used to determine accident consequences for the Spent Nuclear Fuel Project safety documents. It does not determine accident consequences or describe specific accident scenarios, but instead provides generic information.

Preliminary Concept of Operations for the Spent Fuel Management System--WM2017

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cumberland, Riley M; Adeniyi, Abiodun Idowu; Howard, Rob L

The Nuclear Fuels Storage and Transportation Planning Project (NFST) within the U.S. Department of Energy s Office of Nuclear Energy is tasked with identifying, planning, and conducting activities to lay the groundwork for developing interim storage and transportation capabilities in support of an integrated waste management system. The system will provide interim storage for commercial spent nuclear fuel (SNF) from reactor sites and deliver it to a repository. The system will also include multiple subsystems, potentially including; one or more interim storage facilities (ISF); one or more repositories; facilities to package and/or repackage SNF; and transportation systems. The project teammore » is analyzing options for an integrated waste management system. To support analysis, the project team has developed a Concept of Operations document that describes both the potential integrated system and inter-dependencies between system components. The goal of this work is to aid systems analysts in the development of consistent models across the project, which involves multiple investigators. The Concept of Operations document will be updated periodically as new developments emerge. At a high level, SNF is expected to travel from reactors to a repository. SNF is first unloaded from reactors and placed in spent fuel pools for wet storage at utility sites. After the SNF has cooled enough to satisfy loading limits, it is placed in a container at reactor sites for storage and/or transportation. After transportation requirements are met, the SNF is transported to an ISF to store the SNF until a repository is developed or directly to a repository if available. While the high level operation of the system is straightforward, analysts must evaluate numerous alternative options. Alternative options include the number of ISFs (if any), ISF design, the stage at which SNF repackaging occurs (if any), repackaging technology, the types of containers used, repository design

Radioactivity of spent TRIGA fuel

NASA Astrophysics Data System (ADS)

Usang, M. D.; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

2015-04-01

Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

Feasible variants for intermediate storage of the spent fuel obtained at NPP Cernavoda, Romania

DOE Office of Scientific and Technical Information (OSTI.GOV)

Radu, M.; Popescu, G.

1993-12-31

The 5 CANDU-PHW Reactors of 600 Standard type of Cernavoda Nuclear Power Plant are under construction and the first unit is expected to be commissioned in 1995, group 2 following after 2 years, and then groups 3, 4 and 5 one each year. In this study there are presented feasible variants for intermediate storage of spent fuel, obtained during 30 years of operation from the stations at Cernavoda. From the solutions applied worldwide, both dry and wet storage have been taken into account. In any of the two variants, a unique intermediate storage will be provided and the storage buildingmore » was proposed to be built in 4 different stages. As a first estimation, considering the fact that, by now Romania has only one nuclear plant of CANDU fuel type the dry variant seems to be the best.« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-03

... breaches.'' Zircaloy is a type of zirconium alloy which includes both Zircaloy-2 and Zircaloy-4 cladding, but does not include M5 cladding. The M5 is a different type of zirconium alloy, which does not... ``zirconium alloy'' clad spent fuel assemblies in the 24PHB DSC, which would include both the ``zircaloy clad...

CUTS FOR MTR EXCAVATION ILLUSTRATE SEDIMENTARY MANTLE OF SOIL AND ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

CUTS FOR MTR EXCAVATION ILLUSTRATE SEDIMENTARY MANTLE OF SOIL AND GRAVEL OVERLAYING LAVA ROCK FIFTY FEET BELOW. SAGEBRUSH HAS BEEN SCOURED FROM REST OF SITE. CAMERA PROBABLY FACES SOUTHWEST. INL NEGATIVE NO. 67. Unknown Photographer, 6/4/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

75 FR 23821 - Final License Renewal Interim Staff Guidance LR-ISG-2009-01: Aging Management of Spent Fuel Pool...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-04

... approach acceptable to the NRC staff for meeting the requirements of 10 CFR part 54. On December 1, 2009... nuclear power plant spent fuel pool neutron-absorbing materials for compliance with part 54... Regulations (10 CFR part 54). The final LR-ISG revises the NRC staff's aging management recommendations...

Process and apparatus for recovery of fissionable materials from spent reactor fuel by anodic dissolution

DOEpatents

Tomczuk, Zygmunt; Miller, William E.; Wolson, Raymond D.; Gay, Eddie C.

1991-01-01

An electrochemical process and apparatus for the recovery of uranium and plutonium from spent metal clad fuel pins is disclosed. The process uses secondary reactions between U.sup.+4 cations and elemental uranium at the anode to increase reaction rates and improve anodic efficiency compared to prior art processes. In another embodiment of the process, secondary reactions between Cd.sup.+2 cations and elemental uranium to form uranium cations and elemental cadmium also assists in oxidizing the uranium at the anode.

High burn-up spent nuclear fuel transport reliability investigation

DOE PAGES

Wang, Jy-An; Wang, Hong; Jiang, Hao; ...

2018-04-15

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During road or rail transportation, SNF will experience unique conditions that could affect the structural integrity of the cladding due to vibrational and impact loading. Lack of SNF inertia-induced dynamic fatigue data, especially for the high burn-up (HBU) SNF systems, has brought significant challenges to quantify the reliability of SNF during transportation with a high degree of confidence. To address this shortcoming, Oak Ridge National Laboratory (ORNL) developed a SNF vibration testing protocol without fuel pellets removal, which hasmore » provided significant insight regarding the dynamics of mechanical interactions between pellet and cladding. This research has provided a detailed understanding about the effect of loading rate and loading mode on the fatigue damage evolution of HBU SNF under normal conditions of transport (NCT). Static and dynamic loading experimental data were generated for SNF under simulated transportation environments using a cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at ORNL. SNF flexural tensile strength and fatigue S-N data from pressurized water reactors (PWRs) and boiling water reactor (BWR) HBU SNF are presented in this paper, including the potential effects of pellet-cladding interface bonding, hydride reorientation, and thermal annealing to SNF vibration reliability. The data presented here can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in safety of SNF transportation operations.« less

High burn-up spent nuclear fuel transport reliability investigation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An; Wang, Hong; Jiang, Hao

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During road or rail transportation, SNF will experience unique conditions that could affect the structural integrity of the cladding due to vibrational and impact loading. Lack of SNF inertia-induced dynamic fatigue data, especially for the high burn-up (HBU) SNF systems, has brought significant challenges to quantify the reliability of SNF during transportation with a high degree of confidence. To address this shortcoming, Oak Ridge National Laboratory (ORNL) developed a SNF vibration testing protocol without fuel pellets removal, which hasmore » provided significant insight regarding the dynamics of mechanical interactions between pellet and cladding. This research has provided a detailed understanding about the effect of loading rate and loading mode on the fatigue damage evolution of HBU SNF under normal conditions of transport (NCT). Static and dynamic loading experimental data were generated for SNF under simulated transportation environments using a cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at ORNL. SNF flexural tensile strength and fatigue S-N data from pressurized water reactors (PWRs) and boiling water reactor (BWR) HBU SNF are presented in this paper, including the potential effects of pellet-cladding interface bonding, hydride reorientation, and thermal annealing to SNF vibration reliability. The data presented here can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in safety of SNF transportation operations.« less

Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hardin, Ernest; Matteo, Edward N.; Hadgu, Teklu

2016-01-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement formore » extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

MTR WING, TRA604, INTERIOR. BASEMENT. WEST CORRIDOR. CAMERA FACES NORTH. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR WING, TRA-604, INTERIOR. BASEMENT. WEST CORRIDOR. CAMERA FACES NORTH. HVAC AREA IS AT RIGHT OF CORRIDOR. INL NEGATIVE NO. HD46-13-3. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Exosome cofactor hMTR4 competes with export adaptor ALYREF to ensure balanced nuclear RNA pools for degradation and export.

PubMed

Fan, Jing; Kuai, Bin; Wu, Guifen; Wu, Xudong; Chi, Binkai; Wang, Lantian; Wang, Ke; Shi, Zhubing; Zhang, Heng; Chen, She; He, Zhisong; Wang, Siyuan; Zhou, Zhaocai; Li, Guohui; Cheng, Hong

2017-10-02

The exosome is a key RNA machine that functions in the degradation of unwanted RNAs. Here, we found that significant fractions of precursors and mature forms of mRNAs and long noncoding RNAs are degraded by the nuclear exosome in normal human cells. Exosome-mediated degradation of these RNAs requires its cofactor hMTR4. Significantly, hMTR4 plays a key role in specifically recruiting the exosome to its targets. Furthermore, we provide several lines of evidence indicating that hMTR4 executes this role by directly competing with the mRNA export adaptor ALYREF for associating with ARS2, a component of the cap-binding complex (CBC), and this competition is critical for determining whether an RNA is degraded or exported to the cytoplasm. Together, our results indicate that the competition between hMTR4 and ALYREF determines exosome recruitment and functions in creating balanced nuclear RNA pools for degradation and export. © 2017 The Authors.

ORIGAMI Automator Primer. Automated ORIGEN Source Terms and Spent Fuel Storage Pool Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wieselquist, William A.; Thompson, Adam B.; Bowman, Stephen M.

2016-04-01

Source terms and spent nuclear fuel (SNF) storage pool decay heat load analyses for operating nuclear power plants require a large number of Oak Ridge Isotope Generation and Depletion (ORIGEN) calculations. SNF source term calculations also require a significant amount of bookkeeping to track quantities such as core and assembly operating histories, spent fuel pool (SFP) residence times, heavy metal masses, and enrichments. The ORIGEN Assembly Isotopics (ORIGAMI) module in the SCALE code system provides a simple scheme for entering these data. However, given the large scope of the analysis, extensive scripting is necessary to convert formats and process datamore » to create thousands of ORIGAMI input files (one per assembly) and to process the results into formats readily usable by follow-on analysis tools. This primer describes a project within the SCALE Fulcrum graphical user interface (GUI) called ORIGAMI Automator that was developed to automate the scripting and bookkeeping in large-scale source term analyses. The ORIGAMI Automator enables the analyst to (1) easily create, view, and edit the reactor site and assembly information, (2) automatically create and run ORIGAMI inputs, and (3) analyze the results from ORIGAMI. ORIGAMI Automator uses the standard ORIGEN binary concentrations files produced by ORIGAMI, with concentrations available at all time points in each assembly’s life. The GUI plots results such as mass, concentration, activity, and decay heat using a powerful new ORIGEN Post-Processing Utility for SCALE (OPUS) GUI component. This document includes a description and user guide for the GUI, a step-by-step tutorial for a simplified scenario, and appendices that document the file structures used.« less

Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel – design concept and experimental demonstration

DOE PAGES

Henzlova, Daniela; Menlove, Howard Olsen; Rael, Carlos D.; ...

2015-10-09

Our paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. We describe the initial feasibility demonstration of the CIPN instrument, which involved measurements of fourmore » pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. Lastly, these features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.« less

Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel – design concept and experimental demonstration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henzlova, Daniela; Menlove, Howard Olsen; Rael, Carlos D.

Our paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. We describe the initial feasibility demonstration of the CIPN instrument, which involved measurements of fourmore » pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. Lastly, these features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.« less

U.S. sent fuel shipment experience by rail

DOE Office of Scientific and Technical Information (OSTI.GOV)

Colborn, K.

2007-07-01

As planning for the large scale shipment of spent nuclear fuel to Yucca Mountain proceeds to address these challenges, actual shipments of spent fuel in other venues continues to provide proof that domestic rail spent fuel shipments can proceed safely and effectively. This paper presents some examples of recently completed spent fuel shipments, and the shipment of large low-level radioactive waste shipments offering lessons learned that may be beneficial to the planning process for large scale spent fuel shipments in the US. (authors)