Consolidated fuel reprocessing program

NASA Astrophysics Data System (ADS)

1985-02-01

Improved processes and components for the Breeder Reprocessing Engineering Test (BRET) were identified and developed as well as the design, procurement and development of prototypic equipment. The integrated testing of process equipment and flowsheets prototypical of a pilot scale full reprocessing plant, and also for testing prototypical remote features of specific complex components in the system are provided. Information to guide the long range activities of the Consolidated Fuel Reprocessing Program (CERP), a focal point for foreign exchange activities, and support in specialized technical areas are described. Research and development activities in HTGR fuel treatment technology are being conducted. Head-end process and laboratory scale development efforts, as well as studies specific to HTGR fuel, are reported. The development of off-gas treatment processes has generic application to fuel reprocessing, progress in this work is also reported.

Consolidated fuel reprocessing program

NASA Astrophysics Data System (ADS)

1985-04-01

A survey of electrochemical methods applications in fuel reprocessing was completed. A dummy fuel assembly shroud was cut using the remotely operated laser disassembly equipment. Operations and engineering efforts have continued to correct equipment operating, software, and procedural problems experienced during the previous uranium compaigns. Fuel cycle options were examined for the liquid metal reactor fuel cycle. In high temperature gas cooled reactor spent fuel studies, preconceptual designs were completed for the concrete storage cask and open field drywell storage concept. These and other tasks operating under the consolidated fuel reprocessing program are examined.

Flowsheets and source terms for radioactive waste projections

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

Forsberg, C.W.

1985-03-01

Flowsheets and source terms used to generate radioactive waste projections in the Integrated Data Base (IDB) Program are given. Volumes of each waste type generated per unit product throughput have been determined for the following facilities: uranium mining, UF/sub 6/ conversion, uranium enrichment, fuel fabrication, boiling-water reactors (BWRs), pressurized-water reactors (PWRs), and fuel reprocessing. Source terms for DOE/defense wastes have been developed. Expected wastes from typical decommissioning operations for each facility type have been determined. All wastes are also characterized by isotopic composition at time of generation and by general chemical composition. 70 references, 21 figures, 53 tables.

Evaluation of Methods for Decladding LWR Fuel for a Pyroprocessing-Based Reprocessing Plant

DTIC Science & Technology

1992-10-01

oAD-A275 326 ORN.rFM-1121o04 OAK RIDGE NATIONAL LABORATORY Evaluation of Methods for Decladding _LWR Fuel for a Pyroprocessing -Based Reprocessing...Dist. Category UC-526 EVALUATION OF METHODS FOR DECLADDING LWR FUEL FOR A PYROPROCESSING -BASED REPROCESSING PLANT W. D. Bond J. C. Mailen G. E...decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyroprocesses

An assessment of the attractiveness of material associated with thorium/uranium and uranium closed fuel cycles from a safeguards perspective

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

Bathke, Charles Gary; Wallace, Richard K; Hase, Kevin R

2010-01-01

This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with various proposed nuclear fuel cycles. Specifically, this paper examines two closed fuel cycles. The first fuel cycle examined is a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of plutonium/thorium and {sup 233}U/thorium. The used fuel is then reprocessed using the THOREX process and the actinides are recycled. The second fuel cycle examined consists of conventional light water reactors (LWR) whose fuel is reprocessed for actinides that are then fed to and recycled untilmore » consumed in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). As reprocessing of LWR fuel has already been examined, this paper will focus on the reprocessing of the scheme's fast-spectrum reactors' fuel. This study will indicate what is required to render these materials as having low utility for use in nuclear weapons. Nevertheless, the results of this paper suggest that all reprocessing products evaluated so far need to be rigorously safeguarded and provided high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE). The methodology and key findings will be presented.« less

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

Washiya, Tadahiro; Komaki, Jun; Funasaka, Hideyuki

Japan Atomic Energy Agency (JAEA) has been developing the new aqueous reprocessing system named 'NEXT' (New Extraction system for TRU recovery)1-2, which provides many advantages as waste volume reduction, cost savings by advanced components and simplification of process operation. Advanced head-end systems in the 'NEXT' process consist of fuel disassembly system, fuel shearing system and continuous dissolver system. We developed reliable fuel disassembly system with innovative procedure, and short-length shearing system and continuous dissolver system can be provided highly concentrated dissolution to adapt to the uranium crystallization process. We have carried out experimental studies, and fabrication of engineering-scale test devicesmore » to confirm the systems performance. In this paper, research and development of advanced head-end systems are described. (authors)« less

IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

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

Gilles Youinou; Andrea Alfonsi

2012-03-01

This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis,more » the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.« less

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

Rare Earth Electrochemical Property Measurements and Phase Diagram Development in a Complex Molten Salt Mixture for Molten Salt Recycle

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

Zhang, Jinsuo; Guo, Shaoqiang

Pyroprocessing is a promising alternative for the reprocessing of used nuclear fuel (UNF) that uses electrochemical methods. Compared to the hydrometallurgical reprocessing method, pyroprocessing has many advantages such as reduced volume of radioactive waste, simple waste processing, ability to treat refractory material, and compatibility with fast reactor fuel recycle. The key steps of the process are the electro-refining of the spent metallic fuel in the LiCl-KCl eutectic salt, which can be integrated with an electrolytic reduction step for the reprocessing of spent oxide fuels.

Laser-enhanced chemical reactions and the liquid state. II. Possible applications to nuclear fuel reprocessing

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

DePoorter, G.L.; Rofer-DePoorter, C.K.

1976-01-01

Laser photochemistry is surveyed as a possible improvement upon the Purex process for reprocessing spent nuclear fuel. Most of the components of spent nuclear fuel are photochemically active, and lasers can be used to selectively excite individual chemical species. The great variety of chemical species present and the degree of separation that must be achieved present difficulties in reprocessing. Lasers may be able to improve the necessary separations by photochemical reaction or effects on rates and equilibria of reactions. (auth)

Nuclear Fuel Reprocessing

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

Harold F. McFarlane; Terry Todd

2013-11-01

Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore.more » Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor fuels have been irradiated for different purposes, but the vast majority of commercial fuel is uranium oxide clad in zirconium alloy tubing. As a result, commercial reprocessing plants have relatively narrow technical requirements for used nuclear that is accepted for processing.« less

Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

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

Greenspan, Ehud

2015-11-04

This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the radiation damage constraint of presently verified cladding materials. The S&B core is designed to maximize the fraction of neutrons that radially leak from the seed (or “driver”) into the subcritical blanket and reduce neutron loss via axial leakage. The blanket in the S&B core makes beneficial use of the leaking neutrons for improved economics and resource utilization. A specific objective ofmore » this study is to maximize the fraction of core power that can be generated by the blanket without violating the thermal hydraulic and material constraints. Since the blanket fuel requires no reprocessing along with remote fuel fabrication, a larger fraction of power from the blanket will result in a smaller fuel recycling capacity and lower fuel cycle cost per unit of electricity generated. A unique synergism is found between a low conversion ratio (CR) seed and a B&B blanket fueled by thorium. Among several benefits, this synergism enables the very low leakage S&B cores to have small positive coolant voiding reactivity coefficient and large enough negative Doppler coefficient even when using inert matrix fuel for the seed. The benefits of this synergism are maximized when using an annular seed surrounded by an inner and outer thorium blankets. Among the high-performance S&B cores designed to benefit from this unique synergism are: (1) the ultra-long cycle core that features a cycle length of ~7 years; (2) the high-transmutation rate core where the seed fuel features a TRU CR of 0.0. Its TRU transmutation rate is comparable to that of the reference Advanced Burner Reactor (ABR) with CR of 0.5 and the thorium blanket can generate close to 60% of the core power; but requires only one sixth of the reprocessing and fabrication capacity per unit of core power. Nevertheless, these high-performance cores were designed to set upper bounds on the S&B core performance by using larger height and pressure drop than those of typical SFR design. A study was subsequently undertaken to quantify the tradeoff between S&B core design variables and the core performance. This study concludes that a viable S&B core can be designed without significant deviation from SFR core design practices. For example, the S&B core with 120cm active height will be comparable in volume, HM mass and specific power with the S-PRISM core and could fit within the S-PRISM reactor vessel. 43% of this core power will be generated by the once-through thorium blanket; the required capacity for reprocessing and remote fuel fabrication per unit of electricity generated will be approximately one fifth of that for a comparable ABR. The sodium void worth of this 120cm tall S&B core is significantly less positive than that of the reference ABR and the Doppler coefficient is only slightly smaller even though the seed uses a fertile-free fuel. The seed in the high transmutation core requires inert matrix fuel (TRU-40Zr) that has been successfully irradiated by the Fuel Cycle Research & Development program. An additional sensitivity analysis was later conducted to remove the bias introduced by the discrepancy between radiation damage constraints -- 200 DPA applied for S&B cores and fast fluence of 4x1023 n(>0.1MeV)/cm2 applied for ABR core design. Although the performance characteristics of the S&B cores are sensitive to the radiation damage constraint applied, the S&B cores offer very significant performance improvements relative to the conventional ABR core design when using identical constraint.« less

The used nuclear fuel problem - can reprocessing and consolidated storage be complementary?

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

Phillips, C.; Thomas, I.

2013-07-01

This paper describes our CISF (Consolidated Interim Storage Facilities) and Reprocessing Facility concepts and show how they can be combined with a geologic repository to provide a comprehensive system for dealing with spent fuels in the USA. The performance of the CISF was logistically analyzed under six operational scenarios. A 3-stage plan has been developed to establish the CISF. Stage 1: the construction at the CISF site of only a rail receipt interface and storage pad large enough for the number of casks that will be received. The construction of the CISF Canister Handling Facility, the Storage Cask Fabrication Facility,more » the Cask Maintenance Facility and supporting infrastructure are performed during stage 2. The construction and placement into operation of a water-filled pool repackaging facility is completed for Stage 3. By using this staged approach, the capital cost of the CISF is spread over a number of years. It also allows more time for a final decision on the geologic repository to be made. A recycling facility will be built, this facility will used the NUEX recycling process that is based on the aqueous-based PUREX solvent extraction process, using a solvent of tri-N-butyl phosphate in a kerosene diluent. It is capable of processing spent fuels at a rate of 5 MT per day, at burn-ups up to 50 GWD per ton of spent fuels and a minimum of 5 years out-of-reactor cooling.« less

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

Progress Towards International Repositories

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

McCombie, C.; Chapman, N.

2002-02-27

The nuclear fuel cycle is designed to be very international, with some specialist activities (e.g. fuel fabrication, reprocessing, etc.) being confined to a few countries. Nevertheless, political and public opposition has in the past been faced by proposals to internationalise the back-end of the cycle, in particular waste disposal. Attitudes, however, have been changing recently and there is now more acceptance of the general concept of shared repositories and of specific proposals such as that of Pangea. However, as for national facilities, progress towards implementation of shared repositories will be gradual. Moreover, the best vehicle for promoting the concept maymore » not be a commercial type of organization. Consequently the Pangea project team are currently establishing a widely based Association for this purpose.« less

U.S.-South Korea Relations

DTIC Science & Technology

2013-04-26

reprocessed to make new fuel using a type of reprocessing called pyroprocessing .66 The United States and South Korea are jointly researching pyroprocessing ...solutions to spent fuel disposal. Spent fuel disposal is a key policy issue for South Korean officials, and some see pyroprocessing as a potential solution...proponents of pyroprocessing see it as a way to advance energy independence for South Korea. 66

Epsilon Metal Waste Form for Immobilization of Noble Metals from Used Nuclear Fuel

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

Crum, Jarrod V.; Strachan, Denis M.; Rohatgi, Aashish

2013-10-01

Epsilon metal (ε-metal), an alloy of Mo, Pd, Rh, Ru, and Tc, is being developed as a waste form to treat and immobilize the undissolved solids and dissolved noble metals from aqueous reprocessing of commercial used nuclear fuel. Epsilon metal is an attractive waste form for several reasons: increased durability relative to borosilicate glass, it can be fabricated without additives (100% waste loading), and in addition it also benefits borosilicate glass waste loading by eliminating noble metals from the glass and thus the processing problems related there insolubility in glass. This work focused on the processing aspects of the epsilonmore » metal waste form development. Epsilon metal is comprised of refractory metals resulting in high reaction temperatures to form the alloy, expected to be 1500 - 2000°C making it a non-trivial phase to fabricate by traditional methods. Three commercially available advanced technologies were identified: spark-plasma sintering, microwave sintering, and hot isostatic pressing, and investigated as potential methods to fabricate this waste form. Results of these investigations are reported and compared in terms of bulk density, phase assemblage (X-ray diffraction and elemental analysis), and microstructure (scanning electron microscopy).« less

Epsilon metal waste form for immobilization of noble metals from used nuclear fuel

NASA Astrophysics Data System (ADS)

Crum, Jarrod V.; Strachan, Denis; Rohatgi, Aashish; Zumhoff, Mac

2013-10-01

Epsilon metal (ɛ-metal), an alloy of Mo, Pd, Rh, Ru, and Tc, is being developed as a waste form to treat and immobilize the undissolved solids and dissolved noble metals from aqueous reprocessing of commercial used nuclear fuel. Epsilon metal is an attractive waste form for several reasons: increased durability relative to borosilicate glass, it can be fabricated without additives (100% waste loading), and in addition it also benefits borosilicate glass waste loading by eliminating noble metals from the glass, thus the processing problems related to their insolubility in glass. This work focused on the processing aspects of the epsilon metal waste form development. Epsilon metal is comprised of refractory metals resulting in high alloying temperatures, expected to be 1500-2000 °C, making it a non-trivial phase to fabricate by traditional methods. Three commercially available advanced technologies were identified: spark-plasma sintering, microwave sintering, and hot isostatic pressing, and investigated as potential methods to fabricate this waste form. Results of these investigations are reported and compared in terms of bulk density, phase assemblage (X-ray diffraction and elemental analysis), and microstructure (scanning electron microscopy).

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

NASA Astrophysics Data System (ADS)

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

2016-08-01

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

Availability analysis of an HTGR fuel recycle facility. Summary report

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

Sharmahd, J.N.

1979-11-01

An availability analysis of reprocessing systems in a high-temperature gas-cooled reactor (HTGR) fuel recycle facility was completed. This report summarizes work done to date to define and determine reprocessing system availability for a previously planned HTGR recycle reference facility (HRRF). Schedules and procedures for further work during reprocessing development and for HRRF design and construction are proposed in this report. Probable failure rates, transfer times, and repair times are estimated for major system components. Unscheduled down times are summarized.

DESIGN CRITERIA FOR FUEL DISSOLUTION SYSTEMS AND ASSOCIATED SERVICE FACILITIES. PLANT MODIFICATIONS FOR REPROCESSING NON-PRODUCTION REACTOR FUELS. PROJECT CGC-830

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

Bierman, S.R.; Graf, W.A.; Kass, M.

1960-07-29

Design panameters are presented for phases of the facility to reprocess low-enrichment fuels from nonproduction reactors. Included are plant flowsheets and equipment layouts for fuel element dissolution, centrifugation, solution adjustment, and waste handling. Also included are the basic design criteria for the supporting facilities which service these phases and all other facilites located in the vicinity of the selected building (Bldg. 221-U). (J.R.D.)

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).

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.

Dismantling of the 904 Cell at the HAO/Sud Facility - 13466

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

Vaudey, C.E.; Crosnier, S.; Renouf, M.

2013-07-01

La Hague facility, in France, is the spent fuel recycling plant wherein a part of the fuel coming from some of the French, German, Belgian, Swiss, Dutch and Japanese nuclear reactors is reprocessed before being recycled in order to separate certain radioactive elements. The facility has been successively handled by the CEA (1962-1978), Cogema (1978-2006), and AREVA NC (since 2006). La Hague facility is composed of 3 production units: The UP2-400 production unit started to be operated in 1966 for the reprocessing of UNGG metal fuel. In 1976, following the dropout of the graphite-gas technology by EDF, an HAO workshopmore » to reprocess the fuel from the light water reactors is affiliated and then stopped in 2003. - UP2-400 is partially stopped in 2002 and then definitely the 1 January 2004 and is being dismantled - UP2-800, with the same capacity than UP3, started to be operated in 1994 and is still in operation. And UP3 - UP3 was implemented in 1990 with an annual reprocessing capacity of 800 tons of fuel and is still in operation The combined licensed capacity of UP2-800 and UP3 is 1,700 tons of used fuel. (authors)« less

An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

NASA Astrophysics Data System (ADS)

Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.

2013-10-01

In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

Safety Criticality Standards Using the French CRISTAL Code Package: Application to the AREVA NP UO{sub 2} Fuel Fabrication Plant

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

Doucet, M.; Durant Terrasson, L.; Mouton, J.

2006-07-01

Criticality safety evaluations implement requirements to proof of sufficient sub critical margins outside of the reactor environment for example in fuel fabrication plants. Basic criticality data (i.e., criticality standards) are used in the determination of sub critical margins for all processes involving plutonium or enriched uranium. There are several criticality international standards, e.g., ARH-600, which is one the US nuclear industry relies on. The French Nuclear Safety Authority (DGSNR and its advising body IRSN) has requested AREVA NP to review the criticality standards used for the evaluation of its Low Enriched Uranium fuel fabrication plants with CRISTAL V0, the recentlymore » updated French criticality evaluation package. Criticality safety is a concern for every phase of the fabrication process including UF{sub 6} cylinder storage, UF{sub 6}-UO{sub 2} conversion, powder storage, pelletizing, rod loading, assembly fabrication, and assembly transportation. Until 2003, the accepted criticality standards were based on the French CEA work performed in the late seventies with the APOLLO1 cell/assembly computer code. APOLLO1 is a spectral code, used for evaluating the basic characteristics of fuel assemblies for reactor physics applications, which has been enhanced to perform criticality safety calculations. Throughout the years, CRISTAL, starting with APOLLO1 and MORET 3 (a 3D Monte Carlo code), has been improved to account for the growth of its qualification database and for increasing user requirements. Today, CRISTAL V0 is an up-to-date computational tool incorporating a modern basic microscopic cross section set based on JEF2.2 and the comprehensive APOLLO2 and MORET 4 codes. APOLLO2 is well suited for criticality standards calculations as it includes a sophisticated self shielding approach, a P{sub ij} flux determination, and a 1D transport (S{sub n}) process. CRISTAL V0 is the result of more than five years of development work focusing on theoretical approaches and the implementation of user-friendly graphical interfaces. Due to its comprehensive physical simulation and thanks to its broad qualification database with more than a thousand benchmark/calculation comparisons, CRISTAL V0 provides outstanding and reliable accuracy for criticality evaluations for configurations covering the entire fuel cycle (i.e. from enrichment, pellet/assembly fabrication, transportation, to fuel reprocessing). After a brief description of the calculation scheme and the physics algorithms used in this code package, results for the various fissile media encountered in a UO{sub 2} fuel fabrication plant will be detailed and discussed. (authors)« less

Interim waste storage for the Integral Fast Reactor

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

Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

1991-01-01

The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes thatmore » are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig.« less

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Code of Federal Regulations, 2012 CFR

2012-01-01

... and Related Waste Management Facilities F Appendix F to Part 50 Energy NUCLEAR REGULATORY COMMISSION... Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities 1. Public health... facilities for the temporary storage of highlevel radioactive wastes, may be located on privately owned...

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Code of Federal Regulations, 2013 CFR

2013-01-01

... and Related Waste Management Facilities F Appendix F to Part 50 Energy NUCLEAR REGULATORY COMMISSION... Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities 1. Public health... facilities for the temporary storage of highlevel radioactive wastes, may be located on privately owned...

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Code of Federal Regulations, 2014 CFR

2014-01-01

... and Related Waste Management Facilities F Appendix F to Part 50 Energy NUCLEAR REGULATORY COMMISSION... Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities 1. Public health... facilities for the temporary storage of highlevel radioactive wastes, may be located on privately owned...

Critical review of analytical techniques for safeguarding the thorium-uranium fuel cycle

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

Hakkila, E.A.

1978-10-01

Conventional analytical methods applicable to the determination of thorium, uranium, and plutonium in feed, product, and waste streams from reprocessing thorium-based nuclear reactor fuels are reviewed. Separations methods of interest for these analyses are discussed. Recommendations concerning the applicability of various techniques to reprocessing samples are included. 15 tables, 218 references.

NUCLEAR MATERIAL ATTRACTIVENESS: AN ASSESSMENT OF MATERIAL FROM PHWR'S IN A CLOSED THORIUM FUEL CYCLE

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

Sleaford, B W; Collins, B A; Ebbinghaus, B B

2010-04-26

This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that {sup 233}U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined to date needmore » to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of 'attractiveness levels' that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities. The methodology and key findings will be presented.« less

Nuclear Material Attractiveness: An Assessment of Material from PHWR's in a Closed Thorium Fuel Cycle

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

Sleaford, Brad W.; Ebbinghaus, B. B.; Bradley, Keith S.

2010-06-11

This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies [ , ] that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that 233U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined tomore » date need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of "attractiveness levels" that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities [ ]. The methodology and key findings will be presented.« less

Disposal of spent fuel

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

Blomeke, J O; Ferguson, D E; Croff, A G

1978-01-01

Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed betweenmore » the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom.« less

Using eye movement desensitization and reprocessing to enhance treatment of couples.

PubMed

Protinsky, H; Sparks, J; Flemke, K

2001-04-01

Eye Movement Desensitization and Reprocessing (EMDR) as a clinical technique may enhance treatment effectiveness when applied within a couple therapy approach that is emotionally and experientially oriented. Clinical experience indicates that EMDR-based interventions are useful for accessing, activating, tolerating, and reprocessing the intense emotions that often fuel dysfunctional couple interactions. Using EMDR within conjoint sessions to reprocess negative emotions can amplify intimacy, increase connection, and subsequently lead to a change in problematic relationship patterns.

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

Coble, Jamie; Orton, Christopher; Schwantes, Jon

Abstract—The Multi-Isotope Process (MIP) Monitor provides an efficient approach to monitoring the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of reprocessing streams in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test threemore » fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type. Locally weighted PLS models were fitted on-the-fly to estimate continuous fuel characteristics. Burn up was predicted within 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment within approximately 2% RMSPE. This automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters and material diversions.« less

Japan’s Nuclear Future: Policy Debate, Prospects, and U.S. Interests

DTIC Science & Technology

2008-05-09

raised in particular over the construction of an industrial- scale reprocessing facility in Japan,. Additionally, fast breeder reactors also produce more...Nuclear Fuel Cycle Engineering Laboratories. 10 A fast breeder reactor is a fast neutron reactor that produces more plutonium than it consumes, which can...Japan Nuclear Fuel Limited (JNFL) has built and is currently running active testing on a large - scale commercial reprocessing plant at Rokkasho-mura

Method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions

DOEpatents

Horwitz, E. Philip; Delphin, Walter H.

1979-07-24

A method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions containing these and other values by contacting the waste solution with an extractant of tricaprylmethylammonium nitrate in an inert hydrocarbon diluent which extracts the palladium and technetium values from the waste solution. The palladium and technetium values are recovered from the extractant and from any other coextracted values with a strong nitric acid strip solution.

On the influence of the americium isotopic vector on the cooling time of minor actinides bearing blankets in fast reactors

NASA Astrophysics Data System (ADS)

Kooyman, Timothée; Buiron, Laurent; Rimpault, Gerald

2018-05-01

In the heterogeneous minor actinides transmutation approach, the nuclei to be transmuted are loaded in dedicated targets often located at the core periphery, so that long-lived heavy nuclides are turned into shorter-lived fission products by fission. To compensate for low flux level at the core periphery, the minor actinides content in the targets is set relatively high (around 20 at.%), which has a negative impact on the reprocessing of the targets due to their important decay heat level. After a complete analysis of the main contributors to the heat load of the irradiated targets, it is shown here that the choice of the reprocessing order of the various feeds of americium from the fuel cycle depends on the actual limit for fuel reprocessing. If reprocessing of hot targets is possible, it is more interesting to reprocess first the americium feed with a high 243Am content in order to limit the total cooling time of the targets, while if reprocessing of targets is limited by their decay heat, it is more interesting to wait for an increase in the 241Am content before loading the americium in the core. An optimization of the reprocessing order appears to lead to a decrease of the total cooling time by 15 years compared to a situation where all the americium feeds are mixed together when two feeds from SFR are considered with a high reprocessing limit.

Nuclear fuels - Present and future

NASA Astrophysics Data System (ADS)

Olander, D.

2009-06-01

The important developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of LWR fuels are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H 2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel rod designs, the hydride fuel with liquid-metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the very high temperature reactor and the sodium fast reactor, and the accompanying reprocessing technologies, aqueous-based UREX+1a and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the fuel's behavior under irradiation and in the reprocessing schemes are emphasized.

Accelerator-driven Transmutation of Waste

NASA Astrophysics Data System (ADS)

Venneri, Francesco

1998-04-01

Nuclear waste from commercial power plants contains large quantities of plutonium, other fissionable actinides, and long-lived fission products that are potential proliferation concerns and create challenges for the long-term storage. Different strategies for dealing with nuclear waste are being followed by various countries because of their geologic situations and their views on nuclear energy, reprocessing and non-proliferation. The current United States policy is to store unprocessed spent reactor fuel in a geologic repository. Other countries are opting for treatment of nuclear waste, including partial utilization of the fissile material contained in the spent fuel, prior to geologic storage. Long-term uncertainties are hampering the acceptability and eventual licensing of a geologic repository for nuclear spent fuel in the US, and driving up its cost. The greatest concerns are with the potential for radiation release and exposure from the spent fuel for tens of thousands of years and the possible diversion and use of the actinides contained in the waste for weapons construction. Taking advantage of the recent breakthroughs in accelerator technology and of the natural flexibility of subcritical systems, the Accelerator-driven Transmutation of Waste (ATW) concept offers the United States and other countries the possibility to greatly reduce plutonium, higher actinides and environmentally hazardous fission products from the waste stream destined for permanent storage. ATW does not eliminate the need for, but instead enhances the viability of permanent waste repositories. Far from being limited to waste destruction, the ATW concept also brings to the table new technologies that could be relevant for next-generation power producing reactors. In the ATW concept, spent fuel would be shipped to the ATW site where the plutonium, transuranics and selected long-lived fission products would be destroyed by fission or transmutation in their first and only pass through the facility, using an accelerator-driven subcritical burner cooled by liquid lead/bismuth and limited pyrochemical treatment of the spent fuel and residual waste. This approach contrasts with the present-day practices of aqueous reprocessing (Europe and Japan), in which high purity plutonium is produced and used in the fabrication of fresh mixed oxide fuel (MOX) that is shipped off-site for use in light water reactors.

Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

NASA Astrophysics Data System (ADS)

Andrianova, E. A.; Tsibul'skiy, V. F.

2017-12-01

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and 238U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with 238U alone should be reprocessed in the usual way.

Detection and Monitoring of Airborne Nuclear Waste Materials. Annual Report to Department of Energy.

DTIC Science & Technology

1979-12-04

an active core , its detection by counting techniques is often slow and impractical. For these reasons NRL under contract with DoE undertook develop ...Protection and Measurements, Tritium Measurement Techniques NCRP Report No. 47 (1976). 2. " Development of a Continuous Tritium Monitor for Fuel Reprocessing...Trans. Am. Nucl. Soc. 21, 91 (1975). 146. "Process Behavior of and Environmental Assessments of C Releases from an HTGR Fuel Reprocessing Facility" J. W

Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

NASA Astrophysics Data System (ADS)

Tsibulskiy, V. F.; Andrianova, E. A.; Davidenko, V. D.; Rodionova, E. V.; Tsibulskiy, S. V.

2017-12-01

A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium-plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

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

Youinou, Gilles J.

This article presents the results of a neutronics analysis related to the homogeneous recycling of transuranics (TRU) in PWRs with a MOX fuel using enriched uranium instead of depleted uranium. It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely the use of reprocessed uranium. From a neutronics point of view, it is possible to multi-recycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one third and two thirds of the fleet. Recycling neptunium and americium with plutonium significantlymore » decreases the decay heat of the waste stream between 100 to 1,000 years compared to those of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235uranium and recycling it makes a major difference on the natural uranium needs. For example, a PWR fleet recycling its plutonium, neptunium and americium in MOXEU needs 28 percent more natural uranium than a reference UO 2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19 percent less if the reprocessed uranium is recycled back in the reactors, i.e. a 47 percent difference.« less

Impact of Reprocessed Uranium Management on the Homogeneous Recycling of Transuranics in PWRs

DOE PAGES

Youinou, Gilles J.

2017-05-04

This article presents the results of a neutronics analysis related to the homogeneous recycling of transuranics (TRU) in PWRs with a MOX fuel using enriched uranium instead of depleted uranium. It also addresses an often, if not always, overlooked aspect related to the recycling of TRU in PWRs, namely the use of reprocessed uranium. From a neutronics point of view, it is possible to multi-recycle the entirety of the plutonium with or without neptunium and americium in a PWR fleet using MOX-EU fuel in between one third and two thirds of the fleet. Recycling neptunium and americium with plutonium significantlymore » decreases the decay heat of the waste stream between 100 to 1,000 years compared to those of an open fuel cycle or when only plutonium is recycled. The uranium present in MOX-EU used fuel still contains a significant amount of 235uranium and recycling it makes a major difference on the natural uranium needs. For example, a PWR fleet recycling its plutonium, neptunium and americium in MOXEU needs 28 percent more natural uranium than a reference UO 2 open cycle fleet generating the same energy if the reprocessed uranium is not recycled and 19 percent less if the reprocessed uranium is recycled back in the reactors, i.e. a 47 percent difference.« less

SOUTH ELEVATION OF IRRADIATED FUEL STORAGE FACILITY LOCATED IN FUEL ...

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

SOUTH ELEVATION OF IRRADIATED FUEL STORAGE FACILITY LOCATED IN FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTH. INL PHOTO NUMBER HD-54-15-2. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

NORTH ELEVATION OF IRRADIATED FUEL STORAGE FACILITY LOCATED IN FUEL ...

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

NORTH ELEVATION OF IRRADIATED FUEL STORAGE FACILITY LOCATED IN FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTH. INL PHOTO NUMBER HD-54-16-1. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Evaluation of nuclear fuel reprocessing strategies. 2. LWR fuel storage, recycle economics and plutonium logistics

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

Prince, B.E.; Hadley, S.W.

1983-10-27

This is the second of a two-part report intended as a critical review of certain issues involved with closing the Light Water Reactor (LWR) fuel cycle and establishing the basis for future transition to commercial breeder applications. The report is divided into four main sections consisting of (1) a review of the status of the LWR spent fuel management and storage problem; (2) an analysis of the economic incentives for instituting reprocessing and recycle in LWRs; (3) an analysis of the time-dependent aspects of plutonium economic value particularly as related to the LWR-breeder transition; and (4) an analysis of themore » time-dependent aspects of plutonium requirements and supply relative to this transition.« 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 primarily on storage of SF from commercial operation, the principles described are equally applicable to SF from research and production reactors as well as high-level radioactive waste.« less

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.

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.

Iodine Pathways and Off-Gas Stream Characteristics for Aqueous Reprocessing Plants – A Literature Survey and Assessment

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

R. T. Jubin; D. M. Strachan; N. R. Soelberg

2013-09-01

Used nuclear fuel is currently being reprocessed in only a few countries, notably France, England, Japan, and Russia. The need to control emissions of the gaseous radionuclides to the air during nuclear fuel reprocessing has already been reported for the entire plant. But since the gaseous radionuclides can partition to various different reprocessing off-gas streams, for example, from the head end, dissolver, vessel, cell, and melter, an understanding of each of these streams is critical. These off-gas streams have different flow rates and compositions and could have different gaseous radionuclide control requirements, depending on how the gaseous radionuclides partition. Thismore » report reviews the available literature to summarize specific engineering data on the flow rates, forms of the volatile radionuclides in off-gas streams, distributions of these radionuclides in these streams, and temperatures of these streams. This document contains an extensive bibliography of the information contained in the open literature.« less

EAST ELEVATION OF HIGH BAY ADDITION OF FUEL STORAGE BUILDING ...

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

EAST ELEVATION OF HIGH BAY ADDITION OF FUEL STORAGE BUILDING (CPP-603). INL DRAWING NUMBER 200-0603-00-706-051286. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

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.

CONSTRUCTION PROGRESS PHOTO SHOWING WEST STORAGE BASIN AT FUEL STORAGE ...

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

CONSTRUCTION PROGRESS PHOTO SHOWING WEST STORAGE BASIN AT FUEL STORAGE BUILDING (CPP-603). INL PHOTO NUMBER NRTS-51-689. Unknown Photographer, 1950 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

The benefits of a fast reactor closed fuel cycle in the UK

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

Gregg, R.; Hesketh, K.

2013-07-01

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size,more » so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the fission product will primarily be a function of nuclear energy generated). However, by reprocessing spent fuel, it is possible to immobilise the fission product in a more suitable waste form that has far more superior in-repository performance. (authors)« less

FACILITY LAYOUT OF FUEL STORAGE BUILDING (CPP603) SHOWING STORAGE BASINS, ...

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

FACILITY LAYOUT OF FUEL STORAGE BUILDING (CPP-603) SHOWING STORAGE BASINS, FUEL ELEMENT CUTTING FACILITY, AND DRY GRAPHITE STORAGE FACILITY. INL DRAWING NUMBER 200-0603-00-030-056329. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, 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. The fifth step consists of developing a risk informed methodology to assess the various options available for disposition of LWR spent fuel and to select among them. The sixth step is to modify the current U. S. infrastructure and to create a climate to increase the utilization of uranium and the sustainability of nuclear generated electricity. (author)« less

Experience gained with the Synroc demonstration plant at ANSTO and its relevance to plutonium immobilization

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

Jostsons, A.; Ridal, A.; Mercer, D.J.

1996-05-01

The Synroc Demonstration Plant (SDP) was designed and constructed at Lucas Heights to demonstrate the feasibility of Synroc production on a commercial scale (10 kg/hr) with simulated Purex liquid HLW. Since commissioning of the SDP in 1987, over 6000 kg of Synroc has been fabricated with a range of feeds and waste loadings. The SDP utilises uniaxial hot-pressing to consolidate Synroc. Pressureless sintering and hot-isostatic pressing have also been studied at smaller scales. The results of this extensive process development have been incorporated in a conceptual design for a radioactive plant to condition HLW from a reprocessing plant with amore » capacity to treat 800 tpa of spent LWR fuel. Synroic containing TRU, including Pu, and fission products has been fabricated and characterised in a glove-box facility and hot cells, respectively. The extensive experience in processing of Synroc over the past 15 years is summarised and its relevance to immobilization of surplus plutonium is discussed.« less

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-02

... more sophisticated reprocessing technology. During the Bush Administration, the Global Nuclear Energy... Associated with the Global Nuclear Energy Partnership,'' dated June 27, 2007 (ADAMS ML071800084), directed... on some Global Nuclear Energy Partnership (GNEP) initiatives had waned and it appeared appropriate to...

MicroRaman measurements for nuclear fuel reprocessing applications

DOE PAGES

Casella, Amanda; Lines, Amanda; Nelson, Gilbert; ...

2016-12-01

Treatment and reuse of used nuclear fuel is a key component in closing the nuclear fuel cycle. Solvent extraction reprocessing methods that have been developed contain various steps tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper willmore » focus on the development and application of a new MicroRaman probe for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid in solutions relevant to used nuclear fuel reprocessing. Previous research has successfully demonstrated the applicability on the macroscopic scale, using sample probes requiring larger solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 μL or less, which is a scale reduction of over 1,000 fold in sample size. Finally, this paper will summarize our current work in this area including: comparisons between the macroscopic and microscopic probes for detection limits, optimized channel focusing, and application in a flow cell with varying levels of HNO 3, and UO 2(NO 3) 2.« less

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.

Conceptual designs of NDA instruments for the NRTA system at the Rokkasho Reprocessing Plant

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

Li, T.K.; Klosterbuer, S.F.; Menlove, H.O.

The authors are studying conceptual designs of selected nondestructive assay (NDA) instruments for the near-real-time accounting system at the rokkasho Reprocessing Plant (RRP) of Japan Nuclear Fuel Limited (JNFL). The JNFL RRP is a large-scale commercial reprocessing facility for spent fuel from boiling-water and pressurized-water reactors. The facility comprises two major components: the main process area to separate and produce purified plutonium nitrate and uranyl nitrate from irradiated reactor spent fuels, and the co-denitration process area to combine and convert the plutonium nitrate and uranyl nitrate into mixed oxide (MOX). The selected NDA instruments for conceptual design studies are themore » MOX-product canister counter, holdup measurement systems for calcination and reduction furnaces and for blenders in the co-denitration process, the isotope dilution gamma-ray spectrometer for the spent fuel dissolver solution, and unattended verification systems. For more effective and practical safeguards and material control and accounting at RRP, the authors are also studying the conceptual design for the UO{sub 3} large-barrel counter. This paper discusses the state-of-the-art NDA conceptual design and research and development activities for the above instruments.« less

CONSTRUCTION VIEW FUEL STORAGE BUILDING (CPP603) LOOKING EAST SHOWING ASBESTOS ...

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

CONSTRUCTION VIEW FUEL STORAGE BUILDING (CPP-603) LOOKING EAST SHOWING ASBESTOS SIDING. INL PHOTO NUMBER NRTS-51-1543. Unknown Photographer, 2/28/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO SHOWING FUEL STORAGE BUILDING (CPP603) LOOKING NORTHWEST. ...

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

CONSTRUCTION PROGRESS PHOTO SHOWING FUEL STORAGE BUILDING (CPP-603) LOOKING NORTHWEST. INL PHOTO NUMBER NRTS-50-895. Unknown Photographer, 10/30/1950 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

BUILDING PLANS OF FUEL STORAGE BUILDING (CPP603). INL DRAWING NUMBER ...

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

BUILDING PLANS OF FUEL STORAGE BUILDING (CPP-603). INL DRAWING NUMBER 200-0603-61-299-103029. ALTERNATE ID NUMBER 542-31-B-21. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

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.

PLOT PLAN OF FUEL STORAGE BUILDING (CPP603) SHOWING STORAGE BASINS ...

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

PLOT PLAN OF FUEL STORAGE BUILDING (CPP-603) SHOWING STORAGE BASINS AND PROPOSED LOCATION OF FUEL ELEMENT CUTTING FACILITY. INL DRAWING NUMBER 200-0603-00-706-051287. ALTERNATE ID NUMBER CPP-C-1287. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Diethylene-triamine-penta-acetate administration protocol for radiological emergency medicine in nuclear fuel reprocessing plants.

PubMed

Jin, Yutaka

2008-01-01

Inhalation therapy of diethylene-triamine-penta-acetate (DTPA) should be initiated immediately to workers who have significant incorporation of plutonium, americium or curium in the nuclear fuel reprocessing plant. A newly designed electric mesh nebulizer is a small battery-operated passive vibrating mesh device, in which vibrations in an ultrasonic horn are used to force drug solution through a mesh of micron-sized holes. This nebulizer enables DTPA administration at an early stage in the event of a radiation emergency from contamination from the above radioactive metals.

Survey of simulation methods for modeling pulsed sieve-plate extraction columns

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

Burkhart, L.

1979-03-01

The report first considers briefly the use of liquid-liquid extraction in nuclear fuel reprocessing and then describes the operation of the pulse column. Currently available simulation models of the column are reviewed, and followed by an analysis of the information presently available from which the necessary parameters can be obtained for use in a model of the column. Finally, overall conclusions are given regarding the information needed to develop an accurate model of the column for materials accountability in fuel reprocessing plants. 156 references.

Multivariate analysis of gamma spectra to characterize used nuclear fuel

DOE PAGES

Coble, Jamie; Orton, Christopher; Schwantes, Jon

2017-01-17

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Multivariate analysis of gamma spectra to characterize used nuclear fuel

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

Coble, Jamie; Orton, Christopher; Schwantes, Jon

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Assessment for advanced fuel cycle options in CANDU

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

Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

2013-07-01

The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a drivermore » fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.« less

MISCELLANEOUS SECTIONS AND DETAILS OF FUEL STORAGE BUILDING (CPP603). INL ...

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

MISCELLANEOUS SECTIONS AND DETAILS OF FUEL STORAGE BUILDING (CPP-603). INL DRAWING NUMBER 200-0603-61-299-103032. ALTERNATE ID NUMBER 542-31-B-24. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

WEST ELEVATION OF FUEL STORAGE BUILDING (CPP603). PHOTO TAKEN LOOKING ...

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

WEST ELEVATION OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTHEAST. INL PHOTO NUMBER HD-54-20-1. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

WEST ELEVATIONS AND SECTIONS OF FUEL STORAGE BUILDING (CPP603). INL ...

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

WEST ELEVATIONS AND SECTIONS OF FUEL STORAGE BUILDING (CPP-603). INL DRAWING NUMBER 200-063-61-299-103031. ALTERNATE ID NUMBER 542-31-B-23. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO SHOWING EMPLACEMENT STEEL BEAMS FUEL STORAGE BUILDING ...

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

CONSTRUCTION PROGRESS PHOTO SHOWING EMPLACEMENT STEEL BEAMS FUEL STORAGE BUILDING (CPP-603) LOOKING EAST. INL PHOTO NUMBER NRTS-51-1371. Unknown Photographer, 1/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

SIDING AND ROOF DETAILS OF FUEL STORAGE BUILDING (CPP603). INL ...

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

SIDING AND ROOF DETAILS OF FUEL STORAGE BUILDING (CPP-603). INL DRAWING NUMBER 200-0603-61-299-103033. ALTERNATE ID NUMBER 542-31-B-25. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP603) SHOWING CRANE ASSEMBLY ...

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

INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP-603) SHOWING CRANE ASSEMBLY FOR TRANSFER PIT. INL PHOTO NUMBER NRTS-51-2404. Unknown Photographer, 5/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

PLAN VIEW OF FUEL STORAGE BUILDING (CPP603) SHOWING STORAGE BASINS. ...

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

PLAN VIEW OF FUEL STORAGE BUILDING (CPP-603) SHOWING STORAGE BASINS. INL DRAWING NUMBER 200-0603-00-706-051285. ALTERNATE ID NUMBER CPP-D-1285. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

WEST ELEVATION OF FUEL STORAGE BUILDING (CPP603). PHOTO TAKEN LOOKING ...

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

WEST ELEVATION OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTHEAST. INL PHOTO NUMBER HD-54-20-3. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

SOUTH, EAST, NORTH ELEVATIONS AND SECTIONS OF FUEL STORAGE BUILDING ...

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

SOUTH, EAST, NORTH ELEVATIONS AND SECTIONS OF FUEL STORAGE BUILDING (CPP-603). INL DRAWING NUMBER 200-0603-61-299-103030. ALTERNATE ID NUMBER 542-31-B-22. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR OF SECOND FLOOR CONTROL ROOM OF FUEL STORAGE BUILDING ...

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

INTERIOR OF SECOND FLOOR CONTROL ROOM OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTHWEST. INL PHOTO NUMBER HD-54-19-2. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

NORTHERN PORTION OF WEST ELEVATION OF FUEL STORAGE BUILDING (CPP603). ...

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

NORTHERN PORTION OF WEST ELEVATION OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTHEAST. INL PHOTO NUMBER HD-54-20-4. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

OBLIQUE PHOTO OF NORTHWEST CORNER OF FUEL STORAGE BUILDING (CPP603). ...

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

OBLIQUE PHOTO OF NORTHWEST CORNER OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTHEAST. INL PHOTO NUMBER HD-54-14-4. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

NORTHERN PORTION OF WEST ELEVATION OF FUEL STORAGE BUILDING (CPP603). ...

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

NORTHERN PORTION OF WEST ELEVATION OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTHEAST. INL PHOTO NUMBER HD-54-20-2. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

VIEW OF FECF HOT CELL OF FUEL STORAGE BUILDING (CPP603). ...

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

VIEW OF FECF HOT CELL OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORHTWEST. INL PHOTO NUMBER HD-54-18-3. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

VIEW OF TRANSFER BASIN CORRIDOR OF FUEL STORAGE BUILDING (CPP603). ...

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

VIEW OF TRANSFER BASIN CORRIDOR OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTH. INL PHOTO NUMBER HD-54-17-2. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

OBLIQUE PHOTO OF NORTH ELEVATION OF FUEL STORAGE BUILDING (CPP603). ...

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

OBLIQUE PHOTO OF NORTH ELEVATION OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTH. INL PHOTO NUMBER HD-54-14-3. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

NASA Astrophysics Data System (ADS)

Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

2017-11-01

This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability.

Novel Sorbent Development and Evaluation for the Capture of Krypton and Xenon from Nuclear Fuel Reprocessing Off-Gas Streams

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

Troy G. Garn; Mitchell R. Greenhalgh; Jack D. Law

2013-10-01

The release of volatile radionuclides generated during Used Nuclear Fuel reprocessing in the US will most certainly need to be controlled to meet US regulatory emission limits. A US DOE sponsored Off-Gas Sigma Team has been tasked with a multi-lab collaborative research and development effort to investigate and evaluate emissions and immobilization control technologies for the volatile radioactive species generated from commercial Used Nuclear Fuel (UNF) Reprocessing. Physical Adsorption technology is a simpler and potential economical alternative to cryogenic distillation processes that can be used for the capture of krypton and xenon and has resulted in a novel composite sorbentmore » development procedure using synthesized mordenite as the active material. Utilizing the sorbent development procedure, INL sigma team members have developed two composite sorbents that have been evaluated for krypton and xenon capacities at ambient and 191 K temperature using numerous test gas compositions. Adsorption isotherms have been generated to predict equilibration and maximum capacities enabling modeling to support process equipment scale-up.« less

Novel Sorbent Development and Evaluation for the Capture of Krypton and Xenon from Nuclear Fuel Reprocessing Off-Gas Streams

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

Troy G. Garn; Mitchell R. Greenhalgh; Jack D. Law

2013-09-01

The release of volatile radionuclides generated during Used Nuclear Fuel reprocessing in the US will most certainly need to be controlled to meet US regulatory emission limits. A US DOE sponsored Off-Gas Sigma Team has been tasked with a multi-lab collaborative research and development effort to investigate and evaluate emissions and immobilization control technologies for the volatile radioactive species generated from commercial Used Nuclear Fuel (UNF) Reprocessing. Physical Adsorption technology is a simpler and potential economical alternative to cryogenic distillation processes that can be used for the capture of krypton and xenon and has resulted in a novel composite sorbentmore » development procedure using synthesized mordenite as the active material. Utilizing the sorbent development procedure, INL sigma team members have developed two composite sorbents that have been evaluated for krypton and xenon capacities at ambient and 191 K temperature using numerous test gas compositions. Adsorption isotherms have been generated to predict equilibration and maximum capacities enabling modeling to support process equipment scale-up.« 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.

Reconfigurable and Reprocessable Thermoset Shape Memory Polymer with Synergetic Triple Dynamic Covalent Bonds.

PubMed

Wang, Yongwei; Pan, Yi; Zheng, Zhaohui; Ding, Xiaobin

2018-04-20

Degradable shape memory polymers (SMPs), especially for polyurethane-based SMPs, have shown great potential for biomedical applications. How to reasonably fabricate SMPs with the ideal combination of degradability, shape reconfigurability, and reprocessability is a critical issue and remains a challenge for medical disposable materials. Herein, a shape memory poly(urethane-urea) with synergetic triple dynamic covalent bonds is reported via embedding polycaprolactone unit into poly(urethane-urea) with the hindered urea dynamic bond. The single polymer network is biodegradable, thermadapt, and reprocessable, without sacrificing the outstanding shape memory performance. Such a shape memory network with plasticity and reprocessability is expected to have significant and positive impact on the medical device industry. © 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Nuclear Energy Policy

DTIC Science & Technology

2007-07-12

Nuclear Waste Storage Act of 2007. Requires commercial nuclear power plants to transfer spent fuel from pools to dry storage ...enrichment, spent fuel recycling (also called reprocessing), and other fuel cycle facilities that could be used to produce nuclear weapons materials...that had used the leased fuel , along with supplies of fresh nuclear fuel , according to the GNEP concept; see [http://www.gnep.energy.gov].

Dissolution behavior of MgO based inert matrix fuel for the transmutation of minor actinides

NASA Astrophysics Data System (ADS)

Mühr-Ebert, E. L.; Lichte, E.; Bukaemskiy, A.; Finkeldei, S.; Klinkenberg, M.; Brandt, F.; Bosbach, D.; Modolo, G.

2018-07-01

This study explores the dissolution properties of magnesia-based inert matrix nuclear fuel (IMF) containing transuranium elements (TRU). Pure MgO pellets as well as MgO pellets containing CeO2, as surrogate for TRU oxides, and are considered as model systems for genuine magnesia based inert matrix fuel were fabricated. The aim of this study is to identify conditions at which the matrix material can be selectively dissolved during the head-end reprocessing step, allowing a separation of MgO from the actinides, whereas the actinides remain undissolved. The dissolution behavior was studied in macroscopic batch experiments as a function of nitric acid concentration, dissolution medium volume, temperature, stirring velocity, and pellet density (85, 90, 96, and 99%TD). To mimic pellets with various burn-ups the density of the here fabricated pellets was varied. MgO is soluble even under mild conditions (RT, 2.5 mol/L HNO3). The dissolution rates of MgO at different acid concentrations are rather similar, whereas the dissolution rate is strongly dependent on the temperature. Via a microscopic approach, a model was developed to describe the evolution of the pellet surface area during dissolution and determine a surface normalized dissolution rate. Moreover, dissolution rates of the inert matrix fuel containing CeO2 were determined as a function of the acid concentration and temperature. During the dissolution of MgO/CeO2 pellets the MgO dissolves completely, while CeO2 (>99%) remains undissolved. This study intends to provide a profound understanding of the chemical performance of magnesia based IMF containing fissile material. The feasibility of the dissolution of magnesia based IMF with nitric acid is discussed.

VIEW OF CRANE LOADING AND UNLOADING AREA OF FUEL STORAGE ...

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

VIEW OF CRANE LOADING AND UNLOADING AREA OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTH. INL PHOTO NUMBER HD-54-17-4. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

0BLIQUE PHOTO OF EAST ELEVATION OF FUEL STORAGE BUILDING (CPP603). ...

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

0BLIQUE PHOTO OF EAST ELEVATION OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING WEST. INL PHOTO NUMBER HD-54-15-1. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

VIEW OF CRANE LOADING AND UNLOADING AREA OF FUEL STORAGE ...

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

VIEW OF CRANE LOADING AND UNLOADING AREA OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING SOUTH. INL PHOTO NUMBER HD-54-17-1. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

VIEW OF SOUTH STORAGE BASIN NUMBER 1 OF FUEL STORAGE ...

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

VIEW OF SOUTH STORAGE BASIN NUMBER 1 OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTHEAST. INL PHOTO NUMBER HD-54-18-4. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

VIEW OF MIDDLE STORAGE BASIN NUMBER 2 OF FUEL STORAGE ...

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

VIEW OF MIDDLE STORAGE BASIN NUMBER 2 OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTHEAST. INL PHOTO NUMBER HD-54-17-3. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Process for recovery of palladium from nuclear fuel reprocessing wastes

DOEpatents

Campbell, D.O.; Buxton, S.R.

1980-06-16

Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M; (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound; (c) heating the solution at reflux temperature until precipitation is complete; and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

Process for recovery of palladium from nuclear fuel reprocessing wastes

DOEpatents

Campbell, David O.; Buxton, Samuel R.

1981-01-01

Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M, (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound, (c) heating the solution at reflux temperature until precipitation is complete, and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

Time cycle analysis and simulation of material flow in MOX process layout

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

Chakraborty, S.; Saraswat, A.; Danny, K.M.

The (U,Pu)O{sub 2} MOX fuel is the driver fuel for the upcoming PFBR (Prototype Fast Breeder Reactor). The fuel has around 30% PuO{sub 2}. The presence of high percentages of reprocessed PuO{sub 2} necessitates the design of optimized fuel fabrication process line which will address both production need as well as meet regulatory norms regarding radiological safety criteria. The powder pellet route has highly unbalanced time cycle. This difficulty can be overcome by optimizing process layout in terms of equipment redundancy and scheduling of input powder batches. Different schemes are tested before implementing in the process line with the helpmore » of a software. This software simulates the material movement through the optimized process layout. The different material processing schemes have been devised and validity of the schemes are tested with the software. Schemes in which production batches are meeting at any glove box location are considered invalid. A valid scheme ensures adequate spacing between the production batches and at the same time it meets the production target. This software can be further improved by accurately calculating material movement time through glove box train. One important factor is considering material handling time with automation systems in place.« less

Coupling fuel cycles with repositories: how repository institutional choices may impact fuel cycle design

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

Forsberg, C.; Miller, W.F.

2013-07-01

The historical repository siting strategy in the United States has been a top-down approach driven by federal government decision making but it has been a failure. This policy has led to dispatching fuel cycle facilities in different states. The U.S. government is now considering an alternative repository siting strategy based on voluntary agreements with state governments. If that occurs, state governments become key decision makers. They have different priorities. Those priorities may change the characteristics of the repository and the fuel cycle. State government priorities, when considering hosting a repository, are safety, financial incentives and jobs. It follows that statesmore » will demand that a repository be the center of the back end of the fuel cycle as a condition of hosting it. For example, states will push for collocation of transportation services, safeguards training, and navy/private SNF (Spent Nuclear Fuel) inspection at the repository site. Such activities would more than double local employment relative to what was planned for the Yucca Mountain-type repository. States may demand (1) the right to take future title of the SNF so if recycle became economic the reprocessing plant would be built at the repository site and (2) the right of a certain fraction of the repository capacity for foreign SNF. That would open the future option of leasing of fuel to foreign utilities with disposal of the SNF in the repository but with the state-government condition that the front-end fuel-cycle enrichment and fuel fabrication facilities be located in that state.« less

EAST/WEST TRUCK BAY AREA OF TRANSFER BASIN CORRIDOR OF FUEL ...

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

EAST/WEST TRUCK BAY AREA OF TRANSFER BASIN CORRIDOR OF FUEL STORAGE BUILDING (CPP-603). PHOTO TAKEN LOOKING NORTHWEST. INL PHOTO NUMBER HD-54-19-1. Mike Crane, Photographer, 8/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

DOE PAGES

Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; ...

2017-08-30

Here, this paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li 2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl 2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersivemore » X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability.« less

Potential radiological impact of tornadoes on the safety of Nuclear Fuel Services' West Valley Fuel Reprocessing Plant. 2. Reentrainment and discharge of radioactive materials

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

Davis, W Jr

1981-07-01

This report describes results of a parametric study of quantities of radioactive materials that might be discharged by a tornado-generated depressurization on contaminated process cells within the presently inoperative Nuclear Fuel Services' (NFS) fuel reprocessing facility near West Valley, New York. The study involved the following tasks: determining approximate quantities of radioactive materials in the cells and characterizing particle-size distribution; estimating the degree of mass reentrainment from particle-size distribution and from air speed data presented in Part 1; and estimating the quantities of radioactive material (source term) released from the cells to the atmosphere. The study has shown that improperlymore » sealed manipulator ports in the Process Mechanical Cell (PMC) present the most likely pathway for release of substantial quantities of radioactive material in the atmosphere under tornado accident conditions at the facility.« less

Fuel supply of nuclear power industry with the introduction of fast reactors

NASA Astrophysics Data System (ADS)

Muraviev, E. V.

2014-12-01

The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

Advanced Fuel Cycle Cost Basis – 2017 Edition

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

Dixon, B. W.; Ganda, F.; Williams, K. A.

This report, commissioned by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the DOE Nuclear Technology Research and Development (NTRD) Program (previously the Fuel Cycle Research and Development (FCRD) and the Advanced Fuel Cycle Initiative (AFCI)). The report describes the NTRD cost basis development process, reference information on NTRD cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This reportmore » contains reference cost data for numerous fuel cycle cost modules (modules A-O) as well as cost modules for a number of reactor types (R modules). The fuel cycle cost modules were developed in the areas of natural uranium mining and milling, thorium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, managed decay storage, recycled product storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste. Since its inception, this report has been periodically updated. The last such internal document was published in August 2015 while the last external edition was published in December of 2009 as INL/EXT-07-12107 and is available on the Web at URL: www.inl.gov/technicalpublications/Documents/4536700.pdf. This current report (Sept 2017) is planned to be reviewed for external release, at which time it will replace the 2009 report as an external publication. This information is used in the ongoing evaluation of nuclear fuel cycles by the NE NTRD program.« less

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.

Immobilization of Fast Reactor First Cycle Raffinate

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

Langley, K. F.; Partridge, B. A.; Wise, M.

This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cyclemore » raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.« less

Occupational dose reduction at Department of Energy contractor facilities: Bibliography of selected readings in radiation protection and ALARA

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

Dionne, B.J.; Sullivan, S.G.; Baum, J.W.

1993-12-01

This bibliography contains abstracts relating to various aspects of ALARA program implementation and dose reduction activities, with a focus on DOE facilities. Abstracts included in this bibliography were selected from proceedings of technical meetings, journals, research reports, searches of the DOE Energy, Science and Technology Database (in general, the citation and abstract information is presented as obtained from this database), and reprints of published articles provided by the authors. Facility types and activities covered in the scope of this report include: radioactive waste, uranium enrichment, fuel fabrication, spent fuel storage and reprocessing, facility decommissioning, hot laboratories, tritium production, research, testmore » and production reactors, weapons fabrication and testing, fusion, uranium and plutonium processing, radiography, and aocelerators. Information on improved shielding design, decontamination, containments, robotics, source prevention and control, job planning, improved operational and design techniques, as well as on other topics, has been included. In addition, DOE/EH reports not included in previous volumes of the bibliography are in this volume (abstracts 611 to 684). This volume (Volume 5 of the series) contains 217 abstracts. An author index and a subject index are provided to facilitate use. Both indices contain the abstract numbers from previous volumes, as well as the current volume. Information that the reader feels might be included in the next volume of this bibliography should be submitted to the BNL ALARA Center.« less

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

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

Ragusa, Jean; Vierow, Karen

2011-09-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzedmore » advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.« less

AERIAL SHOWING COMPLETED REMOTE ANALYTICAL FACILITY (CPP627) ADJOINING FUEL PROCESSING ...

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

AERIAL SHOWING COMPLETED REMOTE ANALYTICAL FACILITY (CPP-627) ADJOINING FUEL PROCESSING BUILDING AND EXCAVATION FOR HOT PILOT PLANT TO RIGHT (CPP-640). INL PHOTO NUMBER NRTS-60-1221. J. Anderson, Photographer, 3/22/1960 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Supply of enriched uranium for research reactors

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

Mueller, H.

1997-08-01

Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel onmore » December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.« less

Flowsheet Analysis of U-Pu Co-Crystallization Process as a New Reprocessing System

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

Shunji Homma; Jun-ichi Ishii; Jiro Koga

2006-07-01

A new fuel reprocessing system by U-Pu co-crystallization process is proposed and examined by flowsheet analysis. This reprocessing system is based on the fact that hexavalent plutonium in nitric acid solution is co-crystallized with uranyl nitrate, whereas it is not crystallized when uranyl nitrate does not exist in the solution. The system consists of five steps: dissolution of spent fuel, plutonium oxidation, U-Pu co-crystallization as a co-decontamination, re-dissolution of the crystals, and U re-crystallization as a U-Pu separation. The system requires a recycling of the mother liquor from the U-Pu co-crystallization step and the appropriate recycle ratio is determined bymore » flowsheet analysis such that the satisfactory decontamination is achieved. Further flowsheet study using four different compositions of LWR spent fuels demonstrates that the constant ratio of plutonium to uranium in mother liquor from the re-crystallization step is achieved for every composition by controlling the temperature. It is also demonstrated by comparing to the Purex process that the size of the plant based on the proposed system is significantly reduced. (authors)« less

Records of wells, test borings, and some measured geologic sections near the Western New York Nuclear Service Center, Cattaraugus County, New York

USGS Publications Warehouse

Bergeron, M.P.

1985-01-01

The Western New York Nuclear Service Center (WNYNSC) is a 3 ,336-acre tract of land in northern Cattaraugus County, NY, about 30 mi south of Buffalo. In 1963, 247 acres within the WNYNSC was developed for a nuclear-fuel reprocessing plant and ancillary facilities, including (1) a receiving and storage facility to store fuel prior to reprocessing, (2) underground storage tanks for liquid high-level radioactive wastes from fuel reprocessing, (3) a low-level wastewater treatment plant, and (4) two burial grounds for shallow burial of solid radioactive waste. A series of geologic and hydrologic investigations was done as part of the initial development and construction of the facilities by numerous agencies during 1960-62; these produced a large quantity of well data, some of which are difficult to locate or obtain. This report is a compilation of well and boring data collected during this period. The data include records of 236 wells, geologic logs of 145 wells and 167 test borings, and descriptions of 20 measured geologic sections. Two oversized maps show locations of the reported data. (USGS)

Basic repository environmental assessment design basis, Lavender Canyon site

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

Not Available

1988-01-01

This study examines the engineering factors and costs associated with the construction, operation, and decommissioning of a high-level nuclear waste repository in salt in the Paradox Basin in Lavender Canyon, Utah. The study assumes a repository capacity of 36,000 metric tons of heavy metal (MTHM) of unreprocessed spent fuel and 36,000 MTHM of commercial high-level reprocessing waste, along with 7020 canisters of defense high-level reprocessing waste and associated quantities of remote- and contact-handled transuranic waste (TRU). With the exception of TRU, all the waste forms are placed in 300- to 1000-year-life carbon-steel waste packages in a collocated waste handling andmore » packaging facility (WHPF), which is also described. The construction, operation, and decommissioning of the proposed repository is estimated to cost approximately $5.51 billion. Costs include those for the collocated WHPP, engineering, and contingency, but exclude waste form assembly and shipment to the site and waste package fabrication and shipment to the site. These costs reflect the relative average wage rates of the region and the relatively sound nature of the salt at this site. Construction would require an estimated 7.75 years. Engineering factors and costs are not strongly influenced by environmental considerations. 51 refs., 24 figs., 20 tabs.« less

Ion-Exchanged SAPO-34 membranes for Krypton-Xenon Separation: Control of Permeation Properties and Fabrication of Hollow Fiber Membranes

DOE PAGES

Kwon, Yeon Hye; Min, Byunghyun; Yang, Shaowei; ...

2018-01-29

Separation of radioisotope 85Kr from 136Xe is of importance in used nuclear fuel reprocessing. Membrane separation based on zeolite molecular sieves such as chabazite SAPO- 34 is an attractive alternative to energy-intensive cryogenic distillation. We report the synthesis of SAPO-34 membranes with considerably enhanced performance, via thickness reduction based upon control of a steam-assisted vapor-solid conversion technique followed by ion exchange with alkali metal cations. The reduction of membrane thickness leads to a large increase in Kr permeance from 7.5 gas permeation units (GPU) to 26.3 GPU with ideal Kr/Xe selectivities > 20 at 298 K. Cation-exchanged membranes show largemore » (>50%) increases in selectivity at ambient or slight sub-ambient conditions. The adsorption, diffusion, and permeation characteristics of ionexchanged SAPO-34 materials and membranes are investigated in detail, with potassium exchanged SAPO-34 membranes showing particularly attractive performance. Lastly, we then demonstrate the fabrication of selective SAPO-34 membranes on α-alumina hollow fibers.« less

Ion-Exchanged SAPO-34 membranes for Krypton-Xenon Separation: Control of Permeation Properties and Fabrication of Hollow Fiber Membranes

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

Kwon, Yeon Hye; Min, Byunghyun; Yang, Shaowei

Separation of radioisotope 85Kr from 136Xe is of importance in used nuclear fuel reprocessing. Membrane separation based on zeolite molecular sieves such as chabazite SAPO- 34 is an attractive alternative to energy-intensive cryogenic distillation. We report the synthesis of SAPO-34 membranes with considerably enhanced performance, via thickness reduction based upon control of a steam-assisted vapor-solid conversion technique followed by ion exchange with alkali metal cations. The reduction of membrane thickness leads to a large increase in Kr permeance from 7.5 gas permeation units (GPU) to 26.3 GPU with ideal Kr/Xe selectivities > 20 at 298 K. Cation-exchanged membranes show largemore » (>50%) increases in selectivity at ambient or slight sub-ambient conditions. The adsorption, diffusion, and permeation characteristics of ionexchanged SAPO-34 materials and membranes are investigated in detail, with potassium exchanged SAPO-34 membranes showing particularly attractive performance. Lastly, we then demonstrate the fabrication of selective SAPO-34 membranes on α-alumina hollow fibers.« less

Nuclear Fuel Reprocessing: U.S. Policy Development

DTIC Science & Technology

2006-11-29

to the chemical separation of fissionable uranium and plutonium from irradiated nuclear fuel. The World War II-era Manhattan Project developed...created the Atomic Energy Commission (AEC) and transferred production and control of fissionable materials from the Manhattan Project . As the exclusive

Status of the nuclear measurement stations for the process control of spent fuel reprocessing at AREVA NC/La Hague

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

Eleon, Cyrille; Passard, Christian; Hupont, Nicolas

2015-07-01

Nuclear measurements are used at AREVA NC/La Hague for the monitoring of spent fuel reprocessing. The process control is based on gamma-ray spectroscopy, passive neutron counting and active neutron interrogation, and gamma transmission measurements. The main objectives are criticality and safety, online process monitoring, and the determination of the residual fissile mass and activities in the metallic waste remained after fuel shearing and dissolution (empty hulls, grids, end pieces), which are put in radioactive waste drums before compaction. The whole monitoring system is composed of eight measurement stations which will be described in this paper. The main measurement stations no.more » 1, 3 and 7 are needed for criticality control. Before fuel element shearing for dissolution, station no. 1 allows determining the burn-up of the irradiated fuel by gamma-ray spectroscopy with HP Ge (high purity germanium) detectors. The burn-up is correlated to the {sup 137}Cs and {sup 134}Cs gamma emission rates. The fuel maximal mass which can be loaded in one bucket of the dissolver is estimated from the lowest burn-up fraction of the fuel element. Station no. 3 is dedicated to the control of the correct fuel dissolution, which is performed with a {sup 137}Cs gamma ray measurement with a HP Ge detector. Station no. 7 allows estimating the residual fissile mass in the drums filled with the metallic residues, especially in the hulls, from passive neutron counting (spontaneous fission and alpha-n reactions) and active interrogation (fission prompt neutrons induced by a pulsed neutron generator) with proportional {sup 3}He detectors. The measurement stations have been validated for the reprocessing of Uranium Oxide (UOX) fuels with a burn-up rate up to 60 GWd/t. This paper presents a brief overview of the current status of the nuclear measurement stations. (authors)« less

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2008-10-17

safeguards-irrelevant.” The following facilities and activities were not on the separation list: ! 8 indigenous Indian power reactors ! Fast Breeder ...test Reactor (FTBR) and Prototype Fast Breeder Reactors (PFBR) under construction ! Enrichment facilities ! Spent fuel reprocessing facilities (except...potential use in a bomb. In addition, safeguards on enrichment, reprocessing plants, and breeder reactors would support the 2002 U.S. National Strategy to

Nuclear Energy Policy

DTIC Science & Technology

2008-01-28

2007. Requires commercial nuclear power plants to transfer spent fuel from pools to dry storage casks and then convey title to the Secretary of Energy...far more economical options for reducing fossil fuel use .15 (For more on federal incentives and the economics of nuclear power, see CRS Report RL33442...uranium enrichment, spent fuel recycling (also called reprocessing), and other fuel cycle facilities that could be used to produce nuclear weapons

Status of French reactors

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

Ballagny, A.

1997-08-01

The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (exceptmore » if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.« less

The Role of Ceramics in a Resurgent Nuclear Industry

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

Marra, J

2006-02-28

With fuel oil and natural gas prices near record highs and worldwide energy demands increasing at an alarming rate, there is growing interest in revitalization of the nuclear power industry within the United States and across the globe. Ceramic materials have long played a very important part in the commercial nuclear industry with applications throughout the entire fuel cycle; from fuel fabrication to waste stabilization. As the international community begins to look at advanced fuel cycles that minimize waste and increase proliferation resistance, ceramic materials will play an even larger role. Many of the advanced reactor concepts being evaluated operatemore » at high-temperature requiring the use of durable, heat-resistant materials. Ceramic fuels are being investigated for a variety of Generation IV reactor concepts. These include the traditional TRISO-coated particles as well as advanced inert-matrix fuels. In order to minimize wastes and legacy materials, ceramic processes are also being applied to fuel reprocessing operations. Ceramic materials continue to provide a vital contribution in ''closing the fuel cycle'' by stabilization of associated low-level and high-level wastes in highly durable grout, ceramics, and glass. In the next five years, programs that are currently in the conceptual phase will begin laboratory- and engineering-scale demonstrations. This will require production-scale demonstrations of several ceramic technologies from fuel form development to advanced stabilization methods. Within the next five to ten years, these demonstrations will move to even larger scales and will also include radioactive demonstrations of these advanced technologies. These radioactive demonstrations are critical to program success and will require advances in ceramic materials associated with nuclear energy applications.« less

INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP603) LOOKING SOUTHWEST SHOWING ...

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

INTERIOR VIEW OF FUEL STORAGE BUILDING (CPP-603) LOOKING SOUTHWEST SHOWING STORAGE BASIN IN FOREGROUND, TRANSFER CRANE AND UNLOADER TO LEFT OF NORTH SIDE OF HOT CELL. INL PHOTO NUMBER NRTS-58-157. J. Anderson, Photographer, 1/15/1958 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

IMPACTS ON HUMAN HEALTH FROM THE COAL AND NUCLEAR FUEL CYCLES AND OTHER TECHNOLOGIES ASSOCIATED WITH ELECTRIC POWER GENERATION AND TRANSMISSION

EPA Science Inventory

The report evaluates major public health impacts of electric power generation and transmission associated with the nuclear fuel cycle and with coal use. Only existing technology is evaluated. For the nuclear cycle, effects of future use of fuel reprocessing and long-term radioact...

10 CFR 50.54 - Conditions of licenses.

Code of Federal Regulations, 2012 CFR

2012-01-01

...)(1) Each nuclear power plant or fuel reprocessing plant licensee subject to the quality assurance... irradiated fuel. (ff) For licensees of nuclear power plants that have implemented the earthquake engineering... of rated thermal power only if the Commission finds that the state of onsite emergency preparedness...

10 CFR 50.54 - Conditions of licenses.

Code of Federal Regulations, 2013 CFR

2013-01-01

...)(1) Each nuclear power plant or fuel reprocessing plant licensee subject to the quality assurance... irradiated fuel. (ff) For licensees of nuclear power plants that have implemented the earthquake engineering... of rated thermal power only if the Commission finds that the state of onsite emergency preparedness...

Aspects of remote maintenance in an FRG reprocessing plant from the manufacturer's viewpoint

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

Zeitzchel, G.; Tennie, M.; Saal, G.

In April 1986 a consortium led by Kraftwerk Union AG was commissioned by the German society for nuclear fuel reprocessing (DWK) to build the first West German commercial reprocessing plant for spent fuel assemblies. The main result of the planning efforts regarding remote maintenance operations inside the main process building was the introduction of FEMO technology (FEMO is an acronym based on German for remote handling modular technique). According to this technology the two cells in which the actual reprocessing (which is based on the PUREX technique) takes place are provided with frames to accommodate the process components (tanks, pumps,more » agitators, etc.), each frame together with the components which it supports forming one module. The two cells are inaccessible and windowless. For handling operations each cell is equipped with an overhead crane and a crane-like manipulator carrier system (MTS) with power manipulator. Viewing of the operations from outside the cells is made possible by television (TV) cameras installed at the crane, the MTS, and the manipulator. This paper addresses some examples of problems that still need to be solved in connection with FEMO handling. In particular, the need for close cooperation between the equipment operator, the component designer, the process engineer, the planning engineer, and the licensing authorities will be demonstrated.« less

Integrating repositories with fuel cycles: The airport authority model

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

Forsberg, C.

2012-07-01

The organization of the fuel cycle is a legacy of World War II and the cold war. Fuel cycle facilities were developed and deployed without consideration of the waste management implications. This led to the fuel cycle model of a geological repository site with a single owner, a single function (disposal), and no other facilities on site. Recent studies indicate large economic, safety, repository performance, nonproliferation, and institutional incentives to collocate and integrate all back-end facilities. Site functions could include geological disposal of spent nuclear fuel (SNF) with the option for future retrievability, disposal of other wastes, reprocessing with fuelmore » fabrication, radioisotope production, other facilities that generate significant radioactive wastes, SNF inspection (navy and commercial), and related services such as SNF safeguards equipment testing and training. This implies a site with multiple facilities with different owners sharing some facilities and using common facilities - the repository and SNF receiving. This requires a different repository site institutional structure. We propose development of repository site authorities modeled after airport authorities. Airport authorities manage airports with government-owned runways, collocated or shared public and private airline terminals, commercial and federal military facilities, aircraft maintenance bases, and related operations - all enabled and benefiting the high-value runway asset and access to it via taxi ways. With a repository site authority the high value asset is the repository. The SNF and HLW receiving and storage facilities (equivalent to the airport terminal) serve the repository, any future reprocessing plants, and others with needs for access to SNF and other wastes. Non-public special-built roadways and on-site rail lines (equivalent to taxi ways) connect facilities. Airport authorities are typically chartered by state governments and managed by commissions with members appointed by the state governor, county governments, and city governments. This structure (1) enables state and local governments to work together to maximize job and tax benefits to local communities and the state, (2) provides a mechanism to address local concerns such as airport noise, and (3) creates an institutional structure with large incentives to maximize the value of the common asset, the runway. A repository site authority would have a similar structure and be the local interface to any national waste management authority. (authors)« less

CESAR: A Code for Nuclear Fuel and Waste Characterisation

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

Vidal, J.M.; Grouiller, J.P.; Launay, A.

2006-07-01

CESAR (Simplified Evolution Code Applied to Reprocessing) is a depletion code developed through a joint program between CEA and COGEMA. In the late 1980's, the first use of this code dealt with nuclear measurement at the Laboratories of the La Hague reprocessing plant. The use of CESAR was then extended to characterizations of all entrance materials and for characterisation, via tracer, of all produced waste. The code can distinguish more than 100 heavy nuclides, 200 fission products and 100 activation products, and it can characterise both the fuel and the structural material of the fuel. CESAR can also make depletionmore » calculations from 3 months to 1 million years of cooling time. Between 2003-2005, the 5. version of the code was developed. The modifications were related to the harmonisation of the code's nuclear data with the JEF2.2 nuclear data file. This paper describes the code and explains the extensive use of this code at the La Hague reprocessing plant and also for prospective studies. The second part focuses on the modifications of the latest version, and describes the application field and the qualification of the code. Many companies and the IAEA use CESAR today. CESAR offers a Graphical User Interface, which is very user-friendly. (authors)« less

Metal–organic framework with optimally selective xenon adsorption and separation

DOE PAGES

Banerjee, Debasis; Simon, Cory M.; Plonka, Anna M.; ...

2016-06-13

Nuclear energy is considered among the most viable alternatives to our current fossil fuel based energy economy.1 The mass-deployment of nuclear energy as an emissions-free source requires the reprocessing of used nuclear fuel to mitigate the waste.2 One of the major concerns with reprocessing used nuclear fuel is the release of volatile radionuclides such as Xe and Kr. The most mature process for removing these radionuclides is energy- and capital-intensive cryogenic distillation. Alternatively, porous materials such as metal-organic frameworks (MOFs) have demonstrated the ability to selectively adsorb Xe and Kr at ambient conditions.3-8 High-throughput computational screening of large databases ofmore » porous materials has identified a calcium-based nanoporous MOF, SBMOF-1, as the most selective for Xe over Kr.9,10 Here, we affirm this prediction and report that SBMOF-1 exhibits by far the highest Xe adsorption capacity and a remarkable Xe/Kr selectivity under relevant nuclear reprocessing conditions. The exceptional selectivity of SBMOF-1 is attributed to its pore size tailored to Xe and its dense wall of atoms that constructs a binding site with a high affinity for Xe, as evident by single crystal X-ray diffraction and molecular simulation.« less

10 CFR 50.54 - Conditions of licenses.

Code of Federal Regulations, 2014 CFR

2014-01-01

... chapter. (a)(1) Each nuclear power plant or fuel reprocessing plant licensee subject to the quality... irradiated fuel. (ff) For licensees of nuclear power plants that have implemented the earthquake engineering... of rated thermal power only if the Commission finds that the state of onsite emergency preparedness...

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

Spencer, Barry B.; Bruffey, Stephanie H.; Jordan, Jacob A.

US regulations will require the removal of iodine and tritium, along with other volatile and semi-volatile radionuclides, from the off-gas streams of nuclear fuel reprocessing plants. Advanced tritium pretreatment (TPT) is an additional head-end operation that could be incorporated within nuclear fuel reprocessing plants. It utilizes nitrogen dioxide (NOR2R) as an oxidant to convert UOR2R to UR3ROR8R prior to traditional aqueous dissolution. Advanced TPT can result in the quantitative volatilization of both tritium and iodine. Up-front removal of iodine is of significant advantage because otherwise it distributes to several unit operations and the associated off-gas streams. The off-gas streams willmore » then require treatment to comply with US regulations. Advanced TPT is currently under development at Oak Ridge National Laboratory, and a kilogram-scale hot cell demonstration with used nuclear fuel (UNF) is planned for fiscal year (FY) 2018.« less

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

Green, D.W.; Heinrich, R.R.; Graczyk, D.G.

The ACL activities covered IFR fuel reprocessing, corium-concrete interactions, environmental samples, wastes, WIPP support, Advanced Photon Source, H-Tc superconductors, EBWR vessel, soils, illegal drug detection, quality control, etc.

New Concepts for Compact Space Reactor Power Systems for Space Based Radar Applications: A Feasibility Study

DTIC Science & Technology

1989-12-01

SPENT FUEL REPROCESSING COULD ALSO BE EMPLOYED IRRADIATION EXPERIENCE - EXTREMELY LIMITED - JOINT US/UK PROGRAM (ONGOING) - TUI/KFK PROGRAM (CANCELED...only the use of off-the-shelf technologies. For example, conventional fuel technology (uranium dioxide), conventional thermionic conversion...advanced fuel (Americium oxide, A1TI2O3) and advanced thermionic conversion. Concept C involves use of an advanced fuel (Americium oxide, Arri203

Characterization and simulation of soft gamma-ray mirrors for their use with spent fuel rods at reprocessing facilities

DOE PAGES

Ruz, J.; Descalle, M. A.; Alameda, J. B.; ...

2016-05-24

The use of a grazing incidence optic to selectively reflect K-shell fluorescence emission and isotope-specific lines from special nuclear materials is a highly desirable nondestructive analysis method for use in reprocessing fuel environments. Preliminary measurements have been performed, and a simulation suite has been developed to give insight into the design of the x ray optics system as a function of the source emission, multilayer coating characteristics, and general experimental configurations. As a result, the experimental results are compared to the predictions from our simulation toolkit to illustrate the ray-tracing capability and explore the effect of modified optics in futuremore » measurement campaigns.« less

Technetium-99 and strontium-90: Abundance determination at ultratrace sensitivity by AMS as signatures of undeclared nuclear reprocessing activity

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

McAninch, J.E.; Proctor, I.D.

1995-03-01

The purpose of this White Paper is to examine the use of the ultratrace technique Accelerator Mass Spectrometry (AMS) to lower detection limits for {sup 99}Tc and {sup 90}Sr, and to examine the utility of these isotopes as signatures of a convert reprocessing facility. The International Atomic Energy Agency (IAEA) has committed to improving the effectiveness of the IAEA Safeguards System. This is in some degree a result of the discovery in 1991 of an undeclared Iraqi EMIS program. Recommendations from the March 1993 Consultants Group Meeting have resulted in several studies and follow on field trials to identify environmentalmore » signatures from covert nuclear fuel reprocessing activity. In particular, the April, 1993 reports of the Standing Advisory Group on Safeguards Implementation (SAGSI) identified the long-lived radioisotopes Technetium-99 and strontium-90 as two reliable signatures of fuel reprocessing activity. This report also suggested pathways in the chemical processing of irradiated fuel where these elements would be volatilized and potentially released in amounts detectable with ultratrace sensitivity techniques. Based on measured {sup 99}Tc background levels compiled from a variety of sources, it is estimated that AMS can provide 10% measurements of environmental levels of {sup 99}Tc in a few minutes using modestly sized samples: a few grams for soils, plants, or animal tissues; one to several liters for rain or seawater samples; and tens to hundreds of cubic meters for air sampling. Small sample sizes and high sample throughput result in significant increases in feasibility, cost effectiveness, and quality of data for a regional monitoring program. Similar results are expected for {sup 90}Sr.« less

Reconstituted asbestos matrix for fuel cells

NASA Technical Reports Server (NTRS)

Mcbryar, H.

1975-01-01

Method is described for reprocessing commercially available asbestos matrix stock to yield greater porosity and bubble pressure (due to increased surface tension), improved homogeneity, and greater uniformity.

Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

DOEpatents

Lloyd, M.H.

1981-01-09

Method for direct coprocessing of nuclear fuels derived from a product stream of fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

DOEpatents

Lloyd, Milton H.

1983-01-01

Method for direct coprocessing of nuclear fuels derived from a product stream of a fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

10 CFR Appendix I to Part 110 - Illustrative List of Reprocessing Plant Components Under NRC Export Licensing Authority

Code of Federal Regulations, 2010 CFR

2010-01-01

... transuranic elements. Different technical processes can accomplish this separation. However, over the years Purex has become the most commonly used and accepted process. Purex involves the dissolution of... facilities have process functions similar to each other, including: irradiated fuel element chopping, fuel...

National Policy Implications of Storing Nuclear Waste in the Pacific Region,

DTIC Science & Technology

1981-01-01

US Congress, Senate, Committee on Energy and Natural Resources, Pacific Spent Nuclear Fuel Storage , Hearing...selected. 17 One type of shipping cask which has been used to transport spent fuel assemblies to the Nevada Test Site is a leakproof steel cask that can...discussion the following conclusions on the nuclear waste storage issue appear valid. The Reagan decision to reprocess spent fuel has not changed US

Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, task 17208: Final report

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

Amoroso, J. W.; Marra, J. C.

2015-08-26

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics)more » over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).« less

Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208: Final report

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

Amoroso, J. W.; Marra, J. C.

2015-08-26

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics)more » over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).« less

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 fuel can be part of a system that cost-effectively meets the burnup credit needs of a repository. Behind each of these reasons is a regulatory structure with MC&A requirements. In the case of the IAEA, the accountable quantity is elemental plutonium. The material in spent fuel (fissile isotopes, fission products, etc.) emits signatures that provide information about the content and history of the fuel. A variety of nondestructive assay (NDA) techniques are available to quantify these signatures. The effort presented in this paper is investigation of the capabilities of 12 NDA techniques. For these 12, none is conceptually capable of independently determining the Pu content in a spent fuel assembly while at the same time being able to detect the diversion of a significant quantity of rods. For this reason the authors are investigating the capability of 12 NDA techniques with the end goal of integrating a few techniques together into a system that is capable of measuring Pu mass in an assembly. The work described here is the beginning of what is anticipated to be a five year effort: (1) two years of modeling to select the best technologies, (2) one year fabricating instruments and (3) two years measuring spent fuel. This paper describes the first two years of this work. In order to cost effectively and robustly model the performance of the 12 NDA techniques, an 'assembly library' was created. The library contains the following: (a) A diverse range of PWR spent fuel assemblies (burnup, enrichment, cooling time) similar to that which exists in spent pools today and in the future. (b) Diversion scenarios that capture a range of possible rod removal options. (c) The spatial and isotopic detail needed to accurately quantify the capability of all the NDA techniques so as to enable integration. It is our intention to make this library available to other researchers in the field for inter-comparison purposes. The performance of each instrument will be quantified for the full assembly library for measurements in three different media: air, water and borated water. The 12 NDA techniques being researched are the following: Delayed Gamma, Delayed Neutrons, Differential Die-Away, Lead Slowing Down Spectrometer, Neutron Multiplicity, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Passive Neutron Albedo Reactivity, Self-integration Neutron Resonance Densitometry, Total Neutron (Gross Neutron), X-Ray Fluorescence, {sup 252}Cf Interrogation with Prompt Neutron Detection.« less

Development of Crystallizer for Advanced Aqueous Reprocessing Process

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

Tadahiro Washiya; Atsuhiro Shibata; Toshiaki Kikuchi

2006-07-01

Crystallization is one of the remarkable technologies for future fuel reprocessing process that has safety and economical advantages. Japan Atomic Energy Agency (JAEA) (former Japan Nuclear Cycle Development Institute), Mitsubishi Material Corporation and Saitama University have been developing the crystallization process. In previous study, we carried out experimental studies with uranium, MOX and spent fuel conditions, and flowsheet analysis was considered. In association with these studies, an innovative continuous crystallizer and its system was developed to ensure high process performance. From the design study, an annular type continuous crystallizer was selected as the most promising design, and performance was confirmedmore » by small-scale test and engineering scale demonstration at uranium crystallization conditions. In this paper, the design study and the demonstration test results are described. (authors)« less

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

Between-cycle laser system for depressurization and resealing of modified design nuclear fuel assemblies

DOEpatents

Bradley, John G.

1982-01-01

A laser beam is used to puncture fuel cladding for release of contained pressurized fission gas from plenum sections or irradiated fuel pins. Exhausted fission gases are collected and trapped for safe disposal. The laser beam, adjusted to welding mode, is subsequently used to reseal the puncture holes. The fuel assembly is returned to additional irradiation or, if at end of reactivity lifetime, is routed to reprocess. The fuel assembly design provides graded cladding lengths, by rows or arrays, such that the cladding of each component fuel element of the assembly is accessible to laser beam reception.

HLW Return from France to Germany - 15 Years of Experience in Public Acceptance and Technical Aspects - 12149

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

Graf, Wilhelm

Since in 1984 the national reprocessing concept was abandoned the reprocessing abroad was the only existing disposal route until 1994. With the amendment of the Atomic Energy Act in 2001 spent fuel management changed completely since from 1 June 2005 any delivery of spent fuel to reprocessing plants was prohibited and the direct disposal of spent fuel became mandatory. Until 2005 the total amount of spent fuel to be reprocessed abroad added up to 6080 t HM, 5309 t HM thereof in France. The waste generated from reprocessing - alternatively an equivalent amount of radioactive material - has to bemore » returned to the country of origin according to the commercial contracts signed between the German utilities and COGEMA, now AREVA NC, in France and BNFL, now INS in UK. In addition the German and the French government exchanged notes with the obligation of both sides to enable and support the return of reprocessing residues or equivalents to Germany. The return of high active vitrified waste from La Hague to the interim storage facility at Gorleben was demanding from the technical view i. e. the cask design and the transport. Unfortunately the Gorleben area served as a target for nuclear opponents from the first transport in 1996 to the latest one in 2011. The protection against sabotage of the railway lines and mass protests needed highly improved security measures. In France and Germany special working forces and projects have been set up to cope with this extraordinary situation. A complex transport organization was established to involve all parties in line with the German and French requirements during transport. The last transport of vitrified residues from France has been completed successfully so far thus confirming the efficiency of the applied measures. Over 15 years there was and still is worldwide no comparable situation it is still unique. Summing up, the exceptional project handling challenge that resulted from the continuous anti-nuclear civil disobedience in Germany over the whole 15-year long project running time could be faced efficiently. It has to be concluded that despite of all problems the anti-nuclear activities have caused so far, all transports of vitrified HLW have always been completed successfully by adapting the commonly established safety, security and public acceptance measures to the special conditions and needs in Germany and coordinating the activities of all parties involved but at the expense of high costs for industry and government and a challenging operational complexity. Apart from an anticipatory project planning a good communication between all involved industrial parties and the French and the German government was the key to the effective management of such shipments and to minimize the radiological, economic, environmental, public and political impact. The future will show how efficiently the gained experience can be used for further return projects which are to be realized since no reprocessed waste has yet been returned from UK and neither the medium-level nor the low-level radioactive waste has been transferred from France to Germany. (author)« less

The Best Defense: Making Maximum Sense of Minimum Deterrence

DTIC Science & Technology

2011-06-01

uranium fuel cycles and has unmatched experience in the thorium fuel cycle.25 Published sources claim India produces between 20 and 40kg of plutonium...nuclear energy was moderate at best. Pakistan‘s first reactor , which it received from the United States, did not become operational until 1965.4...In 1974 Pakistan signed an agreement with France to supply a reprocessing plant for extracting plutonium from spent fuel from power reactors

Expensive Enrichment

ERIC Educational Resources Information Center

Resnikoff, Marvin

1975-01-01

This article presents an economic analysis of the nuclear fuel reprocessing industry. It indicates that while environmental safety devices have improved the working conditions, they have also added ever-increasing costs to this necessary process. (MA)

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

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

Uranium extraction from TRISO-coated fuel particles using supercritical CO2 containing tri-n-butyl phosphate.

PubMed

Zhu, Liyang; Duan, Wuhua; Xu, Jingming; Zhu, Yongjun

2012-11-30

High-temperature gas-cooled reactors (HTGRs) are advanced nuclear systems that will receive heavy use in the future. It is important to develop spent nuclear fuel reprocessing technologies for HTGR. A new method for recovering uranium from tristructural-isotropic (TRISO-) coated fuel particles with supercritical CO(2) containing tri-n-butyl phosphate (TBP) as a complexing agent was investigated. TRISO-coated fuel particles from HTGR fuel elements were first crushed to expose UO(2) pellet fuel kernels. The crushed TRISO-coated fuel particles were then treated under O(2) stream at 750°C, resulting in a mixture of U(3)O(8) powder and SiC shells. The conversion of U(3)O(8) into solid uranyl nitrate by its reaction with liquid N(2)O(4) in the presence of a small amount of water was carried out. Complete conversion was achieved after 60 min of reaction at 80°C, whereas the SiC shells were not converted by N(2)O(4). Uranyl nitrate in the converted mixture was extracted with supercritical CO(2) containing TBP. The cumulative extraction efficiency was above 98% after 20 min of online extraction at 50°C and 25 MPa, whereas the SiC shells were not extracted by TBP. The results suggest an attractive strategy for reprocessing spent nuclear fuel from HTGR to minimize the generation of secondary radioactive waste. Copyright © 2012 Elsevier B.V. All rights reserved.

Gaseous fuel reactors for power systems

NASA Technical Reports Server (NTRS)

Kendall, J. S.; Rodgers, R. J.

1977-01-01

Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

Converting Maturing Nuclear Sites to Integrated Power Production Islands

DOE PAGES

Solbrig, Charles W.

2011-01-01

Nuclear islands, which are integrated power production sites, could effectively sequester and safeguard the US stockpile of plutonium. A nuclear island, an evolution of the integral fast reactor, utilizes all the Transuranics (Pu plus minor actinides) produced in power production, and it eliminates all spent fuel shipments to and from the site. This latter attribute requires that fuel reprocessing occur on each site and that fast reactors be built on-site to utilize the TRU. All commercial spent fuel shipments could be eliminated by converting all LWR nuclear power sites to nuclear islands. Existing LWR sites have the added advantage ofmore » already possessing a license to produce nuclear power. Each could contribute to an increase in the nuclear power production by adding one or more fast reactors. Both the TRU and the depleted uranium obtained in reprocessing would be used on-site for fast fuel manufacture. Only fission products would be shipped to a repository for storage. The nuclear island concept could be used to alleviate the strain of LWR plant sites currently approaching or exceeding their spent fuel pool storage capacity. Fast reactor breeding ratio could be designed to convert existing sites to all fast reactors, or keep the majority thermal.« less

Back-end of the fuel cycle - Indian scenario

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

Wattal, P.K.

Nuclear power has a key role in meeting the energy demands of India. This can be sustained by ensuring robust technology for the back end of the fuel cycle. Considering the modest indigenous resources of U and a huge Th reserve, India has adopted a three stage Nuclear Power Programme (NPP) based on 'closed fuel cycle' approach. This option on 'Recovery and Recycle' serves twin objectives of ensuring adequate supply of nuclear fuel and also reducing the long term radio-toxicity of the wastes. Reprocessing of the spent fuel by Purex process is currently employed. High Level Liquid Waste (HLW) generatedmore » during reprocessing is vitrified and undergoes interim storage. Back-end technologies are constantly modified to address waste volume minimization and radio-toxicity reduction. Long-term management of HLW in Indian context would involve partitioning of long lived minor actinides and recovery of valuable fission products specifically cesium. Recovery of minor actinides from HLW and its recycle is highly desirable for the sustained growth of India's NPPs. In this context, programme for developing and deploying partitioning technologies on industrial scale is pursued. The partitioned elements could be either transmuted in Fast Reactors (FRs)/Accelerated Driven Systems (ADS) as an integral part of sustainable Indian NPP. (authors)« less

129I in the oceans: origins and applications.

PubMed

Raisbeck, G M; Yiou, F

1999-09-30

The quantity of the long lived (half-life 15.7 million years) radioactive isotope 129I in the pre-nuclear age ocean was approximately 100 kg. Various nuclear related activities, including weapons testing, nuclear fuel reprocessing, Chernobyl and other authorized or non-authorized dumping of radioactive waste have increased the ocean inventory of 129I by more than one order of magnitude. The most important of these sources are the direct marine discharges from the commercial reprocessing facilities at La Hague (France) and Sellafield (UK) which have discharged approximately 1640 kg in the English Channel, and approximately 720 kg in the Irish Sea, respectively. We discuss how this 129I can be used as both a 'pathway' and 'transit time' tracer in the North Atlantic and Arctic oceans, as well as a parameter for distinguishing between reprocessed and non-reprocessed nuclear waste in the ocean, and as a proxy for the transport and dilution of other soluble pollutants input to the North Sea.

Promises and Challenges of Thorium Implementation for Transuranic Transmutation - 13550

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

Franceschini, F.; Lahoda, E.; Wenner, M.

2013-07-01

This paper focuses on the challenges of implementing a thorium fuel cycle for recycle and transmutation of long-lived actinide components from used nuclear fuel. A multi-stage reactor system is proposed; the first stage consists of current UO{sub 2} once-through LWRs supplying transuranic isotopes that are continuously recycled and burned in second stage reactors in either a uranium (U) or thorium (Th) carrier. The second stage reactors considered for the analysis are Reduced Moderation Pressurized Water Reactors (RMPWRs), reconfigured from current PWR core designs, and Fast Reactors (FRs) with a burner core design. While both RMPWRs and FRs can in principlemore » be employed, each reactor and associated technology has pros and cons. FRs have unmatched flexibility and transmutation efficiency. RMPWRs have higher fuel manufacturing and reprocessing requirements, but may represent a cheaper solution and the opportunity for a shorter time to licensing and deployment. All options require substantial developments in manufacturing, due to the high radiation field, and reprocessing, due to the very high actinide recovery ratio to elicit the claimed radiotoxicity reduction. Th reduces the number of transmutation reactors, and is required to enable a viable RMPWR design, but presents additional challenges on manufacturing and reprocessing. The tradeoff between the various options does not make the choice obvious. Moreover, without an overarching supporting policy in place, the costly and challenging technologies required inherently discourage industrialization of any transmutation scheme, regardless of the adoption of U or Th. (authors)« 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.

Aqueous and pyrochemical reprocessing of actinide fuels

NASA Astrophysics Data System (ADS)

Toth, L. Mac; Bond, Walter D.; Avens, Larry R.

1993-02-01

Processing of the nuclear fuel actinides has developed in two independent directions—aqueous processing and pyroprocessing. Similarities in the two processes, their goals, and restraints are indicated in brief parallel descriptions along with distinguishing advantages and areas of future development. It is suggested that from a technical viewpoint, the ultimate process might be a hybrid which incorporates the best steps of each process.

Concentration of 129I in aquatic biota collected from a lake adjacent to the spent nuclear fuel reprocessing plant in Rokkasho, Japan.

PubMed

Ueda, Shinji; Kakiuchi, Hideki; Hasegawa, Hidenao; Kawamura, Hidehisa; Hisamatsu, Shun'ichi

2015-11-01

The spent nuclear fuel reprocessing plant in Rokkasho, Japan, has been undergoing final testing since March 2006. During April 2006-October 2008, that spent fuel was cut and chemically processed, the plant discharged (129)I into the atmosphere and coastal waters. To study (129)I behaviour in brackish Lake Obuchi, which is adjacent to the plant, (129)I concentrations in aquatic biota were measured by accelerator mass spectrometry. Owing to (129)I discharge from the plant, the (129)I concentration in the biota started to rise from the background concentration in 2006 and was high during 2007-08. The (129)I concentration has been rapidly decreasing after the fuel cutting and chemically processing were finished. The (129)I concentration factors in the biota were higher than those reported by IAEA for marine organisms and similar to those reported for freshwater biota. The estimated annual committed effective dose due to ingestion of foods with the maximum (129)I concentration in the biota samples was 2.8 nSv y(-1). © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Potential radiological impact of tornadoes on the safety of Nuclear Fuel Services' West Valley Fuel Reprocessing Plant. Volume I. Tornado effects on head-end cell airflow

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

Holloway, L.J.; Andrae, R.W.

1981-09-01

This report describes results of a parametric study of the impacts of a tornado-generated depressurization on airflow in the contaminated process cells within the presently inoperative Nuclear Fuel Services fuel reprocessing facility near West Valley, NY. The study involved the following tasks: (1) mathematical modeling of installed ventilation and abnormal exhaust pathways from the cells and prediction of tornado-induced airflows in these pathways; (2) mathematical modeling of individual cell flow characteristics and prediction of in-cell velocities induced by flows from step 1; and (3) evaluation of the results of steps 1 and 2 to determine whether any of the pathwaysmore » investigated have the potential for releasing quantities of radioactively contaminated air from the main process cells. The study has concluded that in the event of a tornado strike, certain pathways from the cells have the potential to release radioactive materials of the atmosphere. Determination of the quantities of radioactive material released from the cells through pathways identified in step 3 is presented in Part II of this report.« less

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

Estimation of 85Kr dispersion from the spent nuclear fuel reprocessing plant in Rokkasho, Japan, using an atmospheric dispersion model.

PubMed

Abe, K; Iyogi, T; Kawabata, H; Chiang, J H; Suwa, H; Hisamatsu, S

2015-11-01

The spent nuclear fuel reprocessing plant of Japan Nuclear Fuel Limited (JNFL) located in Rokkasho, Japan, discharged small amounts of (85)Kr into the atmosphere during final tests of the plant with actual spent fuel from 31 March 2006 to October 2008. During this period, the gamma-ray dose rates due to discharged (85)Kr were higher than the background rates measured at the Institute for Environmental Sciences and at seven monitoring stations of the Aomori prefectural government and JNFL. The dispersion of (85)Kr was simulated by means of the fifth-generation Penn State/NCAR Mesoscale Model and the CG-MATHEW/ADPIC models (ver. 5.0) with a vertical terrain-following height coordinate. Although the simulated gamma-ray dose rates due to discharged (85)Kr agreed fairly well with measured rates, the agreement between the estimated monthly mean (85)Kr concentrations and the observed concentrations was poor. Improvement of the vertical flow of air may lead to better estimation of (85)Kr dispersion. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Evaluation of Ruthenium Capture Methods for Tritium Pretreatment Off-Gas Streams

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

Spencer, Barry B.; Jubin, Robert Thomas; Bruffey, Stephanie H.

2017-07-01

In the reprocessing of used nuclear fuel, radioactive elements are released into various plant off-gas streams. While much research and development has focused on the abatement of the volatile nuclides 3H, 14C, 85Kr, and 129I, the potential release of semivolatile isotopes that could also report to the off-gas streams in a reprocessing facility has been examined. Ruthenium (as 106Ru) has been identified as one of the semivolatile nuclides requiring the greatest degree of abatement prior to discharging the plant off-gas to the environment.

Developing a concept for a national used fuel interim storage facility in the United States

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

Lewis, Donald Wayne

2013-07-01

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

Method for cleaning solution used in nuclear fuel reprocessing

DOEpatents

Tallent, O.K.; Crouse, D.J.; Mailen, J.C.

1980-12-17

Nuclear fuel processing solution consisting of tri-n-butyl phosphate and dodecane, with a complex of uranium, plutonium, or zirconium and with a solvent degradation product such as di-n-butyl phosphate therein, is contacted with an aqueous solution of a salt formed from hydrazine and either a dicarboxylic acid or a hydroxycarboxylic acid, thereby removing the aforesaid complex from the processing solution.

Method for cleaning solution used in nuclear fuel reprocessing

DOEpatents

Tallent, Othar K.; Crouse, David J.; Mailen, James C.

1982-01-01

Nuclear fuel processing solution consisting of tri-n-butyl phosphate and dodecane, with a complex of uranium, plutonium, or zirconium and with a solvent degradation product such as di-n-butyl phosphate therein, is contacted with an aqueous solution of a salt formed from hydrazine and either a dicarboxylic acid or a hydroxycarboxylic acid, thereby removing the aforesaid complex from the processing solution.

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

Code of Federal Regulations, 2010 CFR

2010-01-01

... undue risk to the health and safety of the public. This appendix establishes quality assurance...: reactor physics, stress, thermal, hydraulic, and accident analyses; compatibility of materials...

ONDRAF/NIRAS and high-level radioactive waste management in Belgium

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

Decamps, F.

1993-12-31

The National Agency for Radioactive Waste and Enriched Fissile Materials, ONDRAF/NIRAS, is a public body with legal personality in charge of managing all radioactive waste on Belgian territory, regardless of its origin and source. It is also entrusted with tasks related to the management of enriched fissile materials, plutonium containing materials and used or unused nuclear fuel, and with certain aspects of the dismantling of closed down nuclear facilities. High-level radioactive waste management comprises essentially and for the time being the storage of high-level liquid waste produced by the former EUROCHEMIC reprocessing plant and of high-level and very high-level heatmore » producing waste resulting from the reprocessing in France of Belgian spent fuel, as well as research and development (R and D) with regard to geological disposal in clay of this waste type.« less

Development of spent fuel reprocessing process based on selective sulfurization: Study on the Pu, Np and Am sulfurization

NASA Astrophysics Data System (ADS)

Kirishima, Akira; Amano, Yuuki; Nihei, Toshifumi; Mitsugashira, Toshiaki; Sato, Nobuaki

2010-03-01

For the recovery of fissile materials from spent nuclear fuel, we have proposed a novel reprocessing process based on selective sulfurization of fission products (FPs). The key concept of this process is utilization of unique chemical property of carbon disulfide (CS2), i.e., it works as a reductant for U3O8 but works as a sulfurizing agent for minor actinides and lanthanides. Sulfurized FPs and minor actinides (MA) are highly soluble to dilute nitric acid while UO2 and PuO2 are hardly soluble, therefore, FPs and MA can be removed from Uranium and Plutonium matrix by selective dissolution. As a feasibility study of this new concept, the sulfurization behaviours of U, Pu, Np, Am and Eu are investigated in this paper by the thermodynamical calculation, phase analysis of chemical analogue elements and tracer experiments.

Advanced ceramic materials for next-generation nuclear applications

NASA Astrophysics Data System (ADS)

Marra, John

2011-10-01

The nuclear industry is at the eye of a 'perfect storm' with fuel oil and natural gas prices near record highs, worldwide energy demands increasing at an alarming rate, and increased concerns about greenhouse gas (GHG) emissions that have caused many to look negatively at long-term use of fossil fuels. This convergence of factors has led to a growing interest in revitalization of the nuclear power industry within the United States and across the globe. Many are surprised to learn that nuclear power provides approximately 20% of the electrical power in the US and approximately 16% of the world-wide electric power. With the above factors in mind, world-wide over 130 new reactor projects are being considered with approximately 25 new permit applications in the US. Materials have long played a very important role in the nuclear industry with applications throughout the entire fuel cycle; from fuel fabrication to waste stabilization. As the international community begins to look at advanced reactor systems and fuel cycles that minimize waste and increase proliferation resistance, materials will play an even larger role. Many of the advanced reactor concepts being evaluated operate at high-temperature requiring the use of durable, heat-resistant materials. Advanced metallic and ceramic fuels are being investigated for a variety of Generation IV reactor concepts. These include the traditional TRISO-coated particles, advanced alloy fuels for 'deep-burn' applications, as well as advanced inert-matrix fuels. In order to minimize wastes and legacy materials, a number of fuel reprocessing operations are being investigated. Advanced materials continue to provide a vital contribution in 'closing the fuel cycle' by stabilization of associated low-level and high-level wastes in highly durable cements, ceramics, and glasses. Beyond this fission energy application, fusion energy will demand advanced materials capable of withstanding the extreme environments of high-temperature plasma systems. Fusion reactors will likely depend on lithium-based ceramics to produce tritium that fuels the fusion plasma, while high-temperature alloys or ceramics will contain and control the hot plasma. All the while, alloys, ceramics, and ceramic-related processes continue to find applications in the management of wastes and byproducts produced by these processes.

Synthesis and properties of reprocessable sulfonated polyimides cross-linked via acid stimulation for use as proton exchange membranes

NASA Astrophysics Data System (ADS)

Zhang, Boping; Ni, Jiangpeng; Xiang, Xiongzhi; Wang, Lei; Chen, Yongming

2017-01-01

Cross-linked sulfonated polyimides are one of the most promising materials for proton exchange membrane (PEM) applications. However, these cross-linked membranes are difficult to reprocess because they are insoluble. In this study, a series of cross-linkable sulfonated polyimides with flexible pendant alkyl side chains containing trimethoxysilyl groups is successfully synthesized. The cross-linkable polymers are highly soluble in common solvents and can be used to prepare tough and smooth films. Before the cross-linking reaction is complete, the membranes can be reprocessed, and the recovery rate of the prepared films falls within an acceptable range. The cross-linked membranes are obtained rapidly when the cross-linkable membranes are immersed in an acid solution, yielding a cross-linking density of the gel fraction of greater than 90%. The cross-linked membranes exhibit high proton conductivities and tensile strengths under hydrous conditions. Compared with those of pristine membranes, the oxidative and hydrolytic stabilities of the cross-linked membranes are significantly higher. The CSPI-70 membrane shows considerable power density in a direct methanol fuel cell (DMFC) test. All of these results suggest that the prepared cross-linked membranes have great potential for applications in proton exchange membrane fuel cells.

The use of nuclear data in the field of nuclear fuel recycling

NASA Astrophysics Data System (ADS)

Martin, Julie-Fiona; Launay, Agnès; Grassi, Gabriele; Binet, Christophe; Lelandais, Jacques; Lecampion, Erick

2017-09-01

AREVA NC La Hague facility is the first step of the nuclear fuel recycling process implemented in France. The processing of the used fuel is governed by high standards of criticality-safety, and strong expectations on the quality of end-products. From the received used fuel assemblies, the plutonium and the uranium are extracted for further energy production purposes within the years following the reprocessing. Furthermore, the ultimate waste - fission products and minor actinides on the one hand, and hulls and end-pieces on the other hand - is adequately packaged for long term disposal. The used fuel is therefore separated into very different materials, and time scales which come into account may be longer than in some other nuclear fields of activity. Given the variety of the handled nuclear materials, as well as the time scales at stake, the importance given to some radionuclides, and hence to the associated nuclear data, can also be specific to the AREVA NC La Hague plant. A study has thus been led to identify a list of the most important radionuclides for the AREVA NC La Hague plant applications, relying on the running constraints of the facility, and the end-products expectations. The activities at the AREVA NC La Hague plant are presented, and the methodology to extract the most important radionuclides for the reprocessing process is detailed.

The myth of the ``proliferation-resistant'' closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Lyman, Edwin S.

2000-07-01

National nuclear energy programs that engage in reprocessing of spent nuclear fuel (SNF) and the development of "closed" nuclear fuel cycles based on the utilization of plutonium process and store large quantities of weapons-usable nuclear materials in forms vulnerable to diversion or theft by national or subnational groups. Proliferation resistance, an idea dating back at least as far as the International Fuel Cycle Evaluation (INFCE) of the late 1970s, is a loosely defined term referring to processes for chemical separation of SNF that do not extract weapons-usable materials in a purified form.

Recovery of 238PuO2 by Molten Salt Oxidation Processing of 238PuO2 Contaminated Combustibles (Part II)

NASA Astrophysics Data System (ADS)

Remerowski, Mary Lynn; Dozhier, C.; Krenek, K.; VanPelt, C. E.; Reimus, M. A.; Spengler, D.; Matonic, J.; Garcia, L.; Rios, E.; Sandoval, F.; Herman, D.; Hart, R.; Ewing, B.; Lovato, M.; Romero, J. P.

2005-02-01

Pu-238 heat sources are used to fuel radioisotope thermoelectric generators (RTG) used in space missions. The demand for this fuel is increasing, yet there are currently no domestic sources of this material. Much of the fuel is material reprocessed from other sources. One rich source of Pu-238 residual material is that from contaminated combustible materials, such as cheesecloth, ion exchange resins and plastics. From both waste minimization and production efficiency standpoints, the best solution is to recover this material. One way to accomplish separation of the organic component from these residues is a flameless oxidation process using molten salt as the matrix for the breakdown of the organic to carbon dioxide and water. The plutonium is retained in the salt, and can be recovered by dissolution of the carbonate salt in an aqueous solution, leaving the insoluble oxide behind. Further aqueous scrap recovery processing is used to purify the plutonium oxide. Recovery of the plutonium from contaminated combustibles achieves two important goals. First, it increases the inventory of Pu-238 available for heat source fabrication. Second, it is a significant waste minimization process. Because of its thermal activity (0.567 W per gram), combustibles must be packaged for disposition with much lower amounts of Pu-238 per drum than other waste types. Specifically, cheesecloth residues in the form of pyrolyzed ash (for stabilization) are being stored for eventual recovery of the plutonium.

Recovery of 238PuO2 by Molten Salt Oxidation Processing of 238PuO2 Contaminated Combustibles (Part II)

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

Remerowski, Mary Lynn; Dozhier, C.; Krenek, K.

2005-02-06

Pu-238 heat sources are used to fuel radioisotope thermoelectric generators (RTG) used in space missions. The demand for this fuel is increasing, yet there are currently no domestic sources of this material. Much of the fuel is material reprocessed from other sources. One rich source of Pu-238 residual material is that from contaminated combustible materials, such as cheesecloth, ion exchange resins and plastics. From both waste minimization and production efficiency standpoints, the best solution is to recover this material. One way to accomplish separation of the organic component from these residues is a flameless oxidation process using molten salt asmore » the matrix for the breakdown of the organic to carbon dioxide and water. The plutonium is retained in the salt, and can be recovered by dissolution of the carbonate salt in an aqueous solution, leaving the insoluble oxide behind. Further aqueous scrap recovery processing is used to purify the plutonium oxide. Recovery of the plutonium from contaminated combustibles achieves two important goals. First, it increases the inventory of Pu-238 available for heat source fabrication. Second, it is a significant waste minimization process. Because of its thermal activity (0.567 W per gram), combustibles must be packaged for disposition with much lower amounts of Pu-238 per drum than other waste types. Specifically, cheesecloth residues in the form of pyrolyzed ash (for stabilization) are being stored for eventual recovery of the plutonium.« less

78 FR 7816 - Quality Assurance Program Requirements (Operations)

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-04

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0021] Quality Assurance Program Requirements (Operations...), DG-1300, ``Quality Assurance Program Requirements (Operations).'' DATES: Submit comments by April 1... CFR Part 50, Appendix B, ``Quality Assurance Criteria for Nuclear power Plants and Fuel Reprocessing...

CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP601) LOOKING NORTHWEST. INL ...

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

CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTHWEST. INL PHOTO NUMBER NRTS-51-1390. Unknown Photographer, 1/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP627). INL PHOTO ...

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

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP-627). INL PHOTO NUMBER NRTS-54-12124. Unknown Photographer, 9/21/1954 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

DETAILS OF REMOTE ANALYTICAL FACILITY (CPP627). INL DRAWING NUMBER 200062700098105071. ...

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

DETAILS OF REMOTE ANALYTICAL FACILITY (CPP-627). INL DRAWING NUMBER 200-0627-00-098-105071. ALTERNATE ID NUMBER 4272-14-108. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP601) LOOKING EAST. INL ...

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

CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP-601) LOOKING EAST. INL PHOTO NUMBER NRTS-51-1547. Unknown Photographer, 2/28/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Processes for Removal and Immobilization of 14C, 129I, and 85Kr

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

Strachan, Denis M.; Bryan, Samuel A.; Henager, Charles H.

2009-10-05

This is a white paper covering the results of a literature search and preliminary experiments on materials and methods to remove and immobilize gaseous radionuclided that come from the reprocessing of spent nuclear fuel.

Occupational dose reduction at Department of Energy contractor facilities: Bibliography of selected readings in radiation protection and ALARA; Volume 5

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

Dionne, B.J.; Sullivan, S.G.; Baum, J.W.

1994-01-01

Promoting the exchange of information related to implementation of the As Low as Reasonably Achievable (ALARA) philosophy is a continuing objective for the Department of Energy (DOE). This report was prepared by the Brookhaven National Laboratory (BNL) ALARA Center for the DOE Office of Health. It contains the fifth in a series of bibliographies on dose reduction at DOE facilities. The BNL ALARA Center was originally established in 1983 under the sponsorship of the Nuclear Regulatory Commission to monitor dose-reduction research and ALARA activities at nuclear power plants. This effort was expanded in 1988 by the DOE`s Office of Environment,more » Safety and Health, to include DOE nuclear facilities. This bibliography contains abstracts relating to various aspects of ALARA program implementation and dose-reduction activities, with a specific focus on DOE facilities. Abstracts included in this bibliography were selected from proceedings of technical meetings, journals, research reports, searches of the DOE Energy, Science and Technology Database (in general, the citation and abstract information is presented as obtained from this database), and reprints of published articles provided by the authors. Facility types and activities covered in the scope of this report include: radioactive waste, uranium enrichment, fuel fabrication, spent fuel storage and reprocessing, facility decommissioning, hot laboratories, tritium production, research, test and production reactors, weapons fabrication and testing, fusion, uranium and plutonium processing, radiography, and accelerators. Information on improved shielding design, decontamination, containments, robotics, source prevention and control, job planning, improved operational and design techniques, as well as on other topics, has been included. In addition, DOE/EH reports not included in previous volumes of the bibliography are in this volume (abstracts 611 to 684). This volume (Volume 5 of the series) contains 217 abstracts.« less

Fusion Applications and Market Evaluation (FAME) Study

DTIC Science & Technology

1988-02-01

fuel from the breeder. Pyrochemical reprocessing is identified as having the potential for low cost, but needs development . The fast-fission designs... Development Administration, "Alternatives for Man- aging Wastes from Reactors and Post-Fission Operations in the LWR Fuel Cycle," ERDA-76-43 (1976). 5...of the ICF program to produce pulsed radiation for military development applications. X-rays can be converted into UV at about 50% energy efficiency

U.S. and South Korean Cooperation in the World Nuclear Energy Market: Major Policy Considerations

DTIC Science & Technology

2010-01-21

a laboratory-scale research program on reprocessing spent fuel with an advanced pyroprocessing technique. However, the level of consensus over the... pyroprocessing option among government agencies, Korean electric utilities, and the public remains uncertain. The current U.S.-Korea 123 agreement...permission. KAERI’s pyroprocessing technology would partially separate plutonium and uranium from spent fuel, but the United States has not allowed the

JPRS Report, Science & Technology, Japan

DTIC Science & Technology

1987-11-12

Change (4) Future Direction Anyway, it has become almost clear that the effect of power recovery cannot be expected from the insulation of...process spent fuels in greater safety and to recover the uranium or plutonium from spent fuels for effective reapplication. In 1974, the PNC began...constructed to serve as a pilot plant that could be used to establish reprocessing technology for the next practical stage. 32 As for enriched uranium

NUCLEAR MATERIAL ATTRACTIVENESS: AN ASSESSMENT OF MATERIAL ASSOCIATED WITH A CLOSED FUEL CYCLE

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

Bathke, C. G.; Ebbinghaus, B.; Sleaford, Brad W.

2010-06-11

This paper examines the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with the various processing steps required for a closed fuel cycle. This paper combines the results from earlier studies that examined the attractiveness of SNM associated with the processing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR with new results for the final, repeated burning of SNM in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). The results of this paper suggest that all reprocessing products evaluated so farmore » need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of "attractiveness levels" that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities. The methodology and key findings will be presented. Additionally, how these attractiveness levels relate to proliferation resistance (e.g. by increasing impediments to the diversion, theft, or undeclared production of SNM for the purpose of acquiring a nuclear weapon), and how they could be used to help inform policy makers, will be discussed.« less

CONSTRUCTION PROGRESS PHOTO SHOWING EXCAVATION PIT FOR MAIN PROCESSING BUILDING ...

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

CONSTRUCTION PROGRESS PHOTO SHOWING EXCAVATION PIT FOR MAIN PROCESSING BUILDING (CPP-601) LOOKING SOUTH. INL PHOTO NUMBER NRTS-50-693. Unknown Photographer, 1950 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

MISCELLANEOUS ARCHITECTURAL DETAILS OF REMOTE ANALYTICAL FACILITY (CPP627). INL DRAWING ...

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

MISCELLANEOUS ARCHITECTURAL DETAILS OF REMOTE ANALYTICAL FACILITY (CPP-627). INL DRAWING NUMBER 200-0627-00-098-105631. ALTERNATE ID NUMBER 4272-814-134. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO SHOWING MAIN PROCESSING BUILDING (CPP601) LOOKING NORTH. ...

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

CONSTRUCTION PROGRESS PHOTO SHOWING MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTH. INL PHOTO NUMBER NRTS-51-1387. Unknown Photographer, 1/31/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

SOUTH ELEVATION OF HOT PILOT PLANT (CPP640) LOOKING NORTH. INL ...

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

SOUTH ELEVATION OF HOT PILOT PLANT (CPP-640) LOOKING NORTH. INL PHOTO NUMBER HD-22-3-1. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP627). INL PHOTO ...

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

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP-627). INL PHOTO NUMBER NRTS-54-12573. R.G. Larsen, Photographer, 10/20/1954 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP627) SHOWING INITIAL ...

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

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP-627) SHOWING INITIAL EXCAVATION. INL PHOTO NUMBER NRTS-54-10703. Unknown Photographer, 5/21/1954 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

SOUTH ELEVATION OF MAIN PROCESSING BUILDING (CPP601) LOOKING NORTH. INL ...

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

SOUTH ELEVATION OF MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTH. INL PHOTO NUMBER HD-22-5-3. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Nitrogen Trifluoride-Based Fluoride- Volatility Separations Process: Initial Studies

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

McNamara, Bruce K.; Scheele, Randall D.; Casella, Andrew M.

2011-09-28

This document describes the results of our investigations on the potential use of nitrogen trifluoride as the fluorinating and oxidizing agent in fluoride volatility-based used nuclear fuel reprocessing. The conceptual process uses differences in reaction temperatures between nitrogen trifluoride and fuel constituents that produce volatile fluorides to achieve separations and recover valuable constituents. We provide results from our thermodynamic evaluations, thermo-analytical experiments, kinetic models, and provide a preliminary process flowsheet. The evaluations found that nitrogen trifluoride can effectively produce volatile fluorides at different temperatures dependent on the fuel constituent.

Evaluation and development plan of NRTA measurement methods for the Rokkasho Reprocessing Plant

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

Li, T.K.; Hakkila, E.A.; Flosterbuer, S.F.

Near-real-time accounting (NRTA) has been proposed as a safeguards method at the Rokkasho Reprocessing Plant (RRP), a large-scale commercial boiling water and pressurized water reactors spent-fuel reprocessing facility. NRTA for RRP requires material balance closures every month. To develop a more effective and practical NRTA system for RRP, we have evaluated NRTA measurement techniques and systems that might be implemented in both the main process and the co-denitration process areas at RRP to analyze the concentrations of plutonium in solutions and mixed oxide powder. Based on the comparative evaluation, including performance, reliability, design criteria, operation methods, maintenance requirements, and estimatedmore » costs for each possible measurement method, recommendations for development were formulated. This paper discusses the evaluations and reports on the recommendation of the NRTA development plan for potential implementation at RRP.« less

The Military Significance of Small Uranium Enrichment Facilities Fed with Low-Enrichment Uranium (Redacted)

DTIC Science & Technology

1969-12-01

a five-year supply of enriched uranium for reactor fuel . Nevertheless, it seems clear that some foreign enrichment developments are approaching a...produc- tion of fissile material could powerfully influence the assessment of risks and benefits of a nuclear weapons development program . Since... program is likely to include the production of its own relatively pure fissile plutonium. This would involve more rapid cycling and reprocessing of fuel

Method for photochemical reduction of uranyl nitrate by tri-N-butyl phosphate and application of this method to nuclear fuel reprocessing

DOEpatents

De Poorter, Gerald L.; Rofer-De Poorter, Cheryl K.

1978-01-01

Uranyl ion in solution in tri-n-butyl phosphate is readily photochemically reduced to U(IV). The product U(IV) may effectively be used in the Purex process for treating spent nuclear fuels to reduce Pu(IV) to Pu(III). The Pu(III) is readily separated from uranium in solution in the tri-n-butyl phosphate by an aqueous strip.

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

Vidal, Jean-Marc; Eschbach, Romain; Launay, Agnes

CEA and AREVA-NC have developed and used a depletion code named CESAR for 30 years. This user-friendly industrial tool provides fast characterizations for all types of nuclear fuel (PWR / UOX or MOX or reprocess Uranium, BWR / UOX or MOX, MTR and SFR) and the wastes associated. CESAR can evaluate 100 heavy nuclides, 200 fission products and 150 activation products (with Helium and Tritium formation). It can also characterize the structural material of the fuel (Zircalloy, stainless steel, M5 alloy). CESAR provides depletion calculations for any reactor irradiation history and from 3 months to 1 million years of coolingmore » time. CESAR5.3 is based on the latest calculation schemes recommended by the CEA and on an international nuclear data base (JEFF-3.1.1). It is constantly checked against the CEA referenced and qualified depletion code DARWIN. CESAR incorporates the CEA qualification based on the dissolution analyses of fuel rod samples and the 'La Hague' reprocessing plant feedback experience. AREVA-NC uses CESAR intensively at 'La Hague' plant, not only for prospective studies but also for characterizations at different industrial facilities all along the reprocessing process and waste conditioning (near 150 000 calculations per year). CESAR is the reference code for AREVA-NC. CESAR is used directly or indirectly with other software, data bank or special equipment in many parts of the La Hague plants. The great flexibility of CESAR has rapidly interested other projects. CESAR became a 'tool' directly integrated in some other softwares. Finally, coupled with a Graphical User Interface, it can be easily used independently, responding to many needs for prospective studies as a support for nuclear facilities or transport. An English version is available. For the principal isotopes of U and Pu, CESAR5 benefits from the CEA experimental validation for the PWR UOX fuels, up to a burnup of 60 GWd/t and for PWR MOX fuels, up to 45 GWd/t. CESAR version 5.3 uses the CEA reference calculation codes for neutron physics with the JEFF-3.1.1 nuclear data set. (authors)« less

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian... reactor following irradiation, the constituent elements of which have not been separated by reprocessing. ...

MISCELLANEOUS ARCHITECTURAL DETAILS AND SECTIONS OF REMOTE ANALYTICAL FACILITY (CPP627). ...

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

MISCELLANEOUS ARCHITECTURAL DETAILS AND SECTIONS OF REMOTE ANALYTICAL FACILITY (CPP-627). INL DRAWING NUMBER 200-0627-00-098-105632. ALTERNATE ID NUMBER 4272-814-135. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO SHOWING EXCAVATION PIT FOR MAIN PROCESSING BUILDING ...

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

CONSTRUCTION PROGRESS PHOTO SHOWING EXCAVATION PIT FOR MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTHWEST. INL PHOTO NUMBER NRTS-50-885. Unknown Photographer, 10/30/1950 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

SOUTH ELEVATION AND DETAILS OF MAIN PROCESSING BUILDING (CPP601). INL ...

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

SOUTH ELEVATION AND DETAILS OF MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103082. ALTERNATE ID NUMBER 542-12-B-76. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

ARCHITECTURAL WALL SECTIONS OF HOT PILOT PLANT (CPP640). INL DRAWING ...

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

ARCHITECTURAL WALL SECTIONS OF HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-0640-00-279-111682. ALTERNATE ID NUMBER 8952-CPP-640-A-5. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

NORTH AND SOUTH SECTIONS OF REMOTE ANALYTICAL FACILITY (CPP627). INL ...

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

NORTH AND SOUTH SECTIONS OF REMOTE ANALYTICAL FACILITY (CPP-627). INL DRAWING NUMBER 200-0627-00-098-105068. ALTERNATE ID NUMBER 4272-14-105. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

AERIAL VIEW OF MAIN PROCESSING BUILDING SHOWING CONSTRUCTION PROGRESS AND ...

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

AERIAL VIEW OF MAIN PROCESSING BUILDING SHOWING CONSTRUCTION PROGRESS AND EXCAVATION FOR LABORATORY ON LEFT. INL PHOTO NUMBER NRTS-51-1759. Unknown Photographer, 3/28/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

EAST AND WEST ELEVATIONS OF REMOTE ANALYTICAL FACILITY (CPP627). INL ...

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

EAST AND WEST ELEVATIONS OF REMOTE ANALYTICAL FACILITY (CPP-627). INL DRAWING NUMBER 200-0627-00-098-105067. ALTERNATE ID NUMBER 4272-14-104. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

EAST AND WEST ELEVATIONS OF MAIN PROCESSING BUILDING (CPP601). INL ...

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

EAST AND WEST ELEVATIONS OF MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103081. ALTERNATE ID NUMBER 542-11-B-75. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF THE REMOTE ANALYTICAL FACILITY OF SHIELDED GLOVE ...

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

INTERIOR PHOTO OF THE REMOTE ANALYTICAL FACILITY OF SHIELDED GLOVE BOXES IN OPERATING CORRIDOR (CPP-627). INL PHOTO NUMBER NRTS-55-1524. Unknown Photographer, 1955 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

MISCELLANEOUS ARCHITECTURAL DETAILS OF HOT PILOT PLANT (CPP640). INL DRAWING ...

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

MISCELLANEOUS ARCHITECTURAL DETAILS OF HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-640-00-279-111684. ALTERNATE ID NUMBER 8952-CPP-640-A-7. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

BUILDING DETAILS AND SECTIONS OF MAIN PROCESSING BUILDING (CPP601). INL ...

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

BUILDING DETAILS AND SECTIONS OF MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103080. ALTERNATE ID NUMBER 542-11-B-74. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

STRUCTURAL DETAILS AND SECTIONS OF MAIN PROCESSING BUILDING (CPP601). INL ...

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

STRUCTURAL DETAILS AND SECTIONS OF MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103079. ALTERNATE ID NUMBER 542-11-B-73. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP627) SHOWING PLACEMENT ...

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

CONSTRUCTION PROGRESS PHOTO OF REMOTE ANALYTICAL FACILITY (CPP-627) SHOWING PLACEMENT OF PIERS. INL PHOTO NUMBER NRTS-54-11716. Unknown Photographer, 8/20/1954 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Fabrication of fuel pin assemblies, phase 3

NASA Technical Reports Server (NTRS)

Keeton, A. R.; Stemann, L. G.

1972-01-01

Five full size and eight reduced length fuel pins were fabricated for irradiation testing to evaluate design concepts for a fast spectrum lithium cooled compact space power reactor. These assemblies consisted of uranium mononitride fuel pellets encased in a T-111 (Ta-8W-2Hf) clad with a tungsten barrier separating fuel and clad. Fabrication procedures were fully qualified by process development and assembly qualification tests. Detailed specifications and procedures were written for the fabrication and assembly of prototype fuel pins.

Progress on Cleaning Up the Only Commercial Nuclear Fuel Reprocessing Facility to Operate in the United States

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

Jackson, T. J.; MacVean, S. A.; Szlis, K. A.

2002-02-26

This paper describes the progress on cleanup of the West Valley Demonstration Project (WVDP), an environmental management project located south of Buffalo, NY. The WVDP was the site of the only commercial nuclear fuel reprocessing facility to have operated in the United States (1966 to 1972). Former fuel reprocessing operations generated approximately 600,000 gallons of liquid high-level radioactive waste stored in underground tanks. The U.S. Congress passed the WVDP Act in 1980 (WVDP Act) to authorize cleanup of the 220-acre facility. The facility is unique in that it sits on the 3,345-acre Western New York Nuclear Service Center (WNYNSC), whichmore » is owned by New York State through the New York State Energy Research and Development Authority (NYSERDA). The U.S. Department of Energy (DOE) has overall responsibility for the cleanup that is authorized by the WVDP Act, paying 90 percent of the WVDP costs; NYSERDA pays 10 percent. West Valley Nuclear Services Company (WVNSCO) is the management contractor at the WVDP. This paper will provide a description of the many accomplishments at the WVDP, including the pretreatment and near completion of vitrification of all the site's liquid high-level radioactive waste, a demonstration of technologies to characterize the remaining material in the high-level waste tanks, the commencement of decontamination and decommissioning (D&D) activities to place the site in a safe configuration for long-term site management options, and achievement of several technological firsts. It will also include a discussion of the complexities involved in completing the WVDP due to the various agency interests that require integration for future cleanup decisions.« 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.

Recovery of transplutonium elements from nuclear reactor waste

DOEpatents

Campbell, David O.; Buxton, Samuel R.

1977-05-24

A method of separating actinide values from nitric acid waste solutions resulting from reprocessing of irradiated nuclear fuels comprises oxalate precipitation of the major portion of actinide and lanthanide values to provide a trivalent fraction suitable for subsequent actinide/lanthanide partition, exchange of actinide and lanthanide values in the supernate onto a suitable cation exchange resin to provide an intermediate-lived raffinate waste stream substantially free of actinides, and elution of the actinide values from the exchange resin. The eluate is then used to dissolve the trivalent oxalate fraction prior to actinide/lanthanide partition or may be combined with the reprocessing waste stream and recycled.

EAST WEST NORTH ELEVATIONS OF MULTICURIE CELL ARCHITECTURAL DETAILS REMOTE ...

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

EAST WEST NORTH ELEVATIONS OF MULTICURIE CELL ARCHITECTURAL DETAILS REMOTE ANALYTICAL FACILITY (CPP-627). INL DRAWING NUMBER 200-00627-00-706-050245. ALTERNATE ID NUMBER AED-D-245. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

ARCHITECTURAL FLOOR PLAN OF OPERATING AREA HOT PILOT PLANT (CPP640). ...

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

ARCHITECTURAL FLOOR PLAN OF OPERATING AREA HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-0640-00-279-111678. ALTERNATE ID NUMBER 8952-CPP-640-A-1. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO REMOTE ANALYTICAL FACILITY (CPP627) SHOWING EMPLACEMENT OF ...

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

CONSTRUCTION PROGRESS PHOTO REMOTE ANALYTICAL FACILITY (CPP-627) SHOWING EMPLACEMENT OF ROOF SLABS. INL PHOTO NUMBER NRTS-54-13463. R.G. Larsen, Photographer, 12/20/1954 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

PLAN SECTIONS AND DETAILS OF CELL HATCHES MAIN PROCESSING BUILDING ...

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

PLAN SECTIONS AND DETAILS OF CELL HATCHES MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-103256. ALTERNATE ID NUMBER 542-11-F-302. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

ARCHITECTURAL DOOR DETAILS AND SCHEDULE OF HOT PILOT PLANT (CPP640). ...

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

ARCHITECTURAL DOOR DETAILS AND SCHEDULE OF HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-640-00-279-111683. ALTERNATE ID NUMBER 8952-CPP-640-A-6. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

ARCHITECTURAL SECTIONS A, B, C, D, OF HOT PILOT PLANT ...

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

ARCHITECTURAL SECTIONS A, B, C, D, OF HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-0640-00-279-111681. ALTERNATE ID NUMBER 8952-CPP-640-A-5. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410

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

Phillips, Chris; Willis, William; Carter, Robert

2013-07-01

Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for usemore » as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions were agreed. These include an assumed reprocessing capacity of 800 tonnes per year, the requirement to remove as waste forms the volatile fission products carbon-14, iodine-129, krypton-85, tritium and ruthenium-106, the restriction of discharge of any water from the facility unless it meets US Environmental Protection Agency drinking water standards, no intentional blending of wastes to lower their classification, and the requirement for the recovered uranium to be sufficiently free from fission products and neutron-absorbing species to allow it to be re-enriched and recycled as nuclear fuel. The results from this work showed that over 99.9% of the radioactivity in the UNF can be concentrated via reprocessing into a fission-product-containing vitrified product, bottles of compressed krypton storage and a cement grout containing the tritium, that together have a volume of only about one eighth the volume of the original UNF. The other waste forms have larger volumes than the original UNF but contain only the remaining 0.1% of the radioactivity. (authors)« less

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.

LIFE Materials: Thermomechanical Effects Volume 5 - Part I

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

Caro, M; DeMange, P; Marian, J

2009-05-07

Improved fuel performance is a key issue in the current Laser Inertial-Confinement Fusion-Fission Energy (LIFE) engine design. LIFE is a fusion-fission engine composed of a {approx}40-tons fuel blanket surrounding a pulsed fusion neutron source. Fusion neutrons get multiplied and moderated in a Beryllium blanket before penetrating the subcritical fission blanket. The fuel in the blanket is composed of millions of fuel pebbles, and can in principle be burned to over 99% FIMA without refueling or reprocessing. This report contains the following chapters: Chapter A: LIFE Requirements for Materials -- LIFE Fuel; Chapter B: Summary of Existing Knowledge; Chapter C: Identificationmore » of Gaps in Knowledge & Vulnerabilities; and Chapter D: Strategy and Future Work.« less

Facility design, construction, and operation

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

NONE

France has been disposing of low-level radioactive waste (LLW) at the Centre de Stockage de la Manche (CSM) since 1969 and now at the Centre de Stockage de l`Aube (CSA) since 1992. In France, several agencies and companies are involved in the development and implementation of LLW technology. The Commissariat a l`Energie Atomic (CEA), is responsible for research and development of new technologies. The Agence National pour la Gestion des Dechets Radioactifs is the agency responsible for the construction and operation of disposal facilities and for wastes acceptance for these facilities. Compagnie Generale des Matieres Nucleaires provides fuel services, includingmore » uranium enrichment, fuel fabrication, and fuel reprocessing, and is thus one generator of LLW. Societe pour les Techniques Nouvelles is an engineering company responsible for commercializing CEA waste management technology and for engineering and design support for the facilities. Numatec, Inc. is a US company representing these French companies and agencies in the US. In Task 1.1 of Numatec`s contract with Martin Marietta Energy Systems, Numatec provides details on the design, construction and operation of the LLW disposal facilities at CSM and CSA. Lessons learned from operation of CSM and incorporated into the design, construction and operating procedures at CSA are identified and discussed. The process used by the French for identification, selection, and evaluation of disposal technologies is provided. Specifically, the decisionmaking process resulting in the change in disposal facility design for the CSA versus the CSM is discussed. This report provides` all of the basic information in these areas and reflects actual experience to date.« less

Fabrication of U-10 wt.%Zr Metallic Fuel Rodlets for Irradiation Test in BOR-60 Fast Reactor

DOE PAGES

Kim, Ki-Hwan; Kim, Jong-Hwan; Oh, Seok-Jin; ...

2016-01-01

The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr fuel slugs with a diameter of 5.5 mm. Consequently, fuel slugs per melting batch without casting defects were fabricated through the developmentmore » of advanced casting technology and evaluation tests. The optimal GTAW welding conditions were also established through a number of experiments. In addition, a qualification test was carried out to prove the weld quality of the end plug welding of the metallic fuel rodlets. The wire wrapping of metallic fuel rodlets was successfully accomplished for the irradiation test. Thus, PGSFR fuel rodlets have been soundly fabricated for the irradiation test in a BOR-60 fast reactor.« less

WEST ELEVATION OF REMOTE ANALYTICAL FACILITY (CPP627) AND HOT PILOT ...

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

WEST ELEVATION OF REMOTE ANALYTICAL FACILITY (CPP-627) AND HOT PILOT PLANT (CPP-640) LOOKING NORTHEAST. INL PHOTO NUMBER HD-22-2-1. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

ARCHITECTURAL FLOOR PLAN OF PROCESS AND ACCESS AREAS HOT PILOT ...

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

ARCHITECTURAL FLOOR PLAN OF PROCESS AND ACCESS AREAS HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-0640-00-279-111679. ALTERNATE ID NUMBER 8952-CPP-640-A-2. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP640) OVERALL VIEW ...

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

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP-640) OVERALL VIEW LOOKING SOUTHEAST; CONSTRUCTION 34 PERCENT COMPLETE. INL PHOTO NUMBER NRTS-60-3034. Holmes, Photographer, 6/23/1960 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

NORTH ELEVATION OF HOT PILOT PLANT (CPP640) LOOKING SOUTH AFTER ...

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

NORTH ELEVATION OF HOT PILOT PLANT (CPP-640) LOOKING SOUTH AFTER REMOTE ANALYTICAL FACILITY (CPP-627) WAS REMOVED. PHOTO NUMBER HD-54-33-2. Mike Crane, Photographer, 7/2006 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP601) ON THE RIGHT ...

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

CONSTRUCTION VIEW OF MAIN PROCESSING BUILDING (CPP-601) ON THE RIGHT AND LABORATORY (CPP-602) ON THE LEFT. INL PHOTO NUMBER NRTS-51-3373. Unknown Photographer, 9/28/1951 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

ARCHITECTURAL ROOF PLAN AND WESTSOUTHEAST ELEVATIONS OF HOT PILOT PLANT ...

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

ARCHITECTURAL ROOF PLAN AND WEST-SOUTHEAST ELEVATIONS OF HOT PILOT PLANT (CPP-640). INL DRAWING NUMBER 200-0640-00-279-111680. ALTERNATE ID NUMBER 8952-CPP-640-A-3. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Process for the extraction of technetium from uranium

DOEpatents

Gong, Cynthia-May S.; Poineau, Frederic; Czerwinski, Kenneth R.

2010-12-21

A spent fuel reprocessing method contacts an aqueous solution containing Technetium(V) and uranyl with an acidic solution comprising hydroxylamine hydrochloride or acetohydroxamic acid to reduce Tc(V) to Tc(II, and then extracts the uranyl with an organic phase, leaving technetium(II) in aqueous solution.

Electrolysis cell for reprocessing plutonium reactor fuel

DOEpatents

Miller, William E.; Steindler, Martin J.; Burris, Leslie

1986-01-01

An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals, the cell including a metallic pot containing a metallic pool as one anode at a lower level, a fused salt as the electrolyte at an intermediate level and a cathode and an anode basket in spaced-apart positions in the electrolyte with the cathode and anode being retractable to positions above the electrolyte during which spent fuel may be added to the anode basket and the anode basket being extendable into the lower pool to dissolve at least some metallic contaminants, the anode basket containing the spent fuel acting as a second anode when in the electrolyte.

Electrolysis cell for reprocessing plutonium reactor fuel

DOEpatents

Miller, W.E.; Steindler, M.J.; Burris, L.

1985-01-04

An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals is claimed. The cell includes a metallic pot containing a metallic pool as one anode at a lower level, a fused salt as the electrolyte at an intermediate level and a cathode and an anode basket in spaced-apart positions in the electrolyte with the cathode and anode being retractable to positions above the electrolyte during which spent fuel may be added to the anode basket. The anode basket is extendable into the lower pool to dissolve at least some metallic contaminants; the anode basket contains the spent fuel acting as a second anode when in the electrolyte.

Documentation and Analysis of IAEA (International Atomic Energy Agency) Safeguards Implementation at the Exxon Nuclear Fuel Fabrication Plant.

DTIC Science & Technology

1984-10-01

SAFEGUARDS AT SIMILAR FACILTTIES ASEA -ATOM LEU FUEL FABRICATION PLANT IN VASTERAS, SWEDEN..................B-1 APPENDIX C - EFFECTS OF NONMEASUREMENT ERRORS...second visit was to the ASEA -ATOM’s fuel fabrication plant in Vasteras, Sweden. The safeguards specialists for those plants were interviewed by R...Facilities, ASEA -ATOM LEU Fuel Fabrication Plant in Vasteras, Sweden, by V. Andersson of ASEA -ATOM, Vasteras, Sweden and R. Nilson of Exxon Nuclear

Fabrication of (U, Zr) C-fueled/tungsten-clad specimens for irradiation in the Plum Brook Reactor Facility

NASA Technical Reports Server (NTRS)

1972-01-01

Fuel samples, 90UC - 10 ZrC, and chemically vapor deposited tungsten fuel cups were fabricated for the study of the long term dimensional stability and compatibility of the carbide-tungsten fuel-cladding systems under irradiation. These fuel samples and fuel cups were assembled into the fuel pins of two capsules, designated as V-2E and V-2F, for irradiation in NASA Plum Brook Reactor Facility at a fission power density of 172 watts/c.c. and a miximum cladding temperature of 1823 K. Fabrication methods and characteristics of the fuel samples and fuel cups prepared are described.

New approaches for MOX multi-recycling

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

Gain, T.; Bouvier, E.; Grosman, R.

Due to its low fissile content after irradiation, Pu from used MOX fuel is considered by some as not recyclable in LWR (Light Water Reactors). The point of this paper is hence to go back to those statements and provide a new analysis based on AREVA extended experience in the fields of fissile and fertile material management and optimized waste management. This is done using the current US fuel inventory as a case study. MOX Multi-recycling in LWRs is a closed cycle scenario where U and Pu management through reprocessing and recycling leads to a significant reduction of the usedmore » assemblies to be stored. The recycling of Pu in MOX fuel is moreover a way to maintain the self-protection of the Pu-bearing assemblies. With this scenario, Pu content is also reduced repetitively via a multi-recycling of MOX in LWRs. Simultaneously, {sup 238}Pu content decreases. All along this scenario, HLW (High-Level Radioactive Waste) vitrified canisters are produced and planned for deep geological disposal. Contrary to used fuel, HLW vitrified canisters do not contain proliferation materials. Moreover, the reprocessing of used fuel limits the space needed on current interim storage. With MOX multi-recycling in LWR, Pu isotopy needs to be managed carefully all along the scenario. The early introduction of a limited number of SFRs (Sodium Fast Reactors) can therefore be a real asset for the overall system. A few SFRs would be enough to improve the Pu isotopy from used LWR MOX fuel and provide a Pu-isotopy that could be mixed back with multi-recycled Pu from LWRs, hence increasing the Pu multi-recycling potential in LWRs.« less

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP601) ...

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

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP-601) LOOKING SOUTHWEST. PHOTO TAKEN FROM NORTHEAST CORNER. INL PHOTO NUMBER HD-50-4-2. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

FIRST FLOOR PLAN OF REMOTE ANALYTICAL FACILITY (CPP627) SHOWING REMOTE ...

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

FIRST FLOOR PLAN OF REMOTE ANALYTICAL FACILITY (CPP-627) SHOWING REMOTE ANALYTICAL LABORATORY, DECONTAMINATION ROOM, AND MULTICURIE CELL ROOM. INL DRAWING NUMBER 200-0627-00-008-105065. ALTERNATE ID NUMBER 4272-14-102. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP601) ...

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

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP-601) LOOKING NORTH. PHOTO TAKEN FROM SOUTHWEST CORNER. INL PHOTO NUMBER HD-50-1-3. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP640) LOOKING NORTHEAST ...

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

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP-640) LOOKING NORTHEAST SHOWING OVERALL BLOCK EXTERIOR WALLS; CONSTRUCTION 65 PERCENT COMPLETE. INL PHOTO NUMBER NRTS-60-4976. Holmes, Photographer, 9/26/1960 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP640) LOOKING EAST ...

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

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP-640) LOOKING EAST SHOWING EXCAVATION AND FORMING; CONSTRUCTION 6 PERCENT COMPLETE. INL PHOTO NUMBER NRTS-59-4935. J. Anderson, Photographer, 9/21/1959 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF MAIN PROCESSING BUILDING (CPP601) PROCESS MAKEUP AREA ...

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

INTERIOR PHOTO OF MAIN PROCESSING BUILDING (CPP-601) PROCESS MAKEUP AREA LOOKING SOUTH. PHOTO TAKEN FROM CENTER OF WEST WALL. INL PHOTO NUMBER HD-50-1-4. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

NORTH AND WEST ELEVATIONS OF REMOTE ANALYTICAL FACILITY (CPP627) LOOKING ...

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

NORTH AND WEST ELEVATIONS OF REMOTE ANALYTICAL FACILITY (CPP-627) LOOKING SOUTHEAST. HEADEND PLANT (CPP-640) APPEARS IN THE BACKGROUND. INL PHOTO NUMBER HD-22-1-4. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF HOT PILOT PLANT SECOND FLOOR DEPICTING DETAIL ...

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

INTERIOR PHOTO OF HOT PILOT PLANT SECOND FLOOR DEPICTING DETAIL OF SHIELDED CAVE (CPP-640) LOOKING SOUTHWEST. PHOTO TAKEN FROM NORTH. INL PHOTO NUMBER HD-54-40-2. Mike Crane, Photographer, 7/2006 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

PLAN SECTIONS AND ELEVATIONS OF VESSEL SAMPLING STATIONS "P", "Q", ...

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

PLAN SECTIONS AND ELEVATIONS OF VESSEL SAMPLING STATIONS "P", "Q", "S" CELLS MAIN PROCESSING BUILDING (CPP-601). INL DRAWING NUMBER 200-0601-00-291-053694. ALTERNATE ID NUMBER CPP-E-1394. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

EAST ELEVATION OF MAIN PROCESSING BUILDING (CPP601) LOOKING NORTHWEST. MAINTENANCE ...

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

EAST ELEVATION OF MAIN PROCESSING BUILDING (CPP-601) LOOKING NORTHWEST. MAINTENANCE SHOP AND OFFICE BUILDING (CPP-630) ON RIGHT IN PHOTO. INL PHOTO NUMBER HD-22-3-2. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF HOT PILOT PLANT SECOND FLOOR WITH SOUTH ...

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

INTERIOR PHOTO OF HOT PILOT PLANT SECOND FLOOR WITH SOUTH SECTION OF SHIELDED CAVE IN FOREGROUND (CPP-640) LOOKING NORTHWEST. INL PHOTO NUMBER HD-54-40-1. Mike Crane, Photographer, 7/2006 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

EQUIPMENT LAYOUT OF MAIN PROCESSING BUILDING (CPP601) LCELL PLAN AND ...

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

EQUIPMENT LAYOUT OF MAIN PROCESSING BUILDING (CPP-601) L-CELL PLAN AND SECTION SHOWS COMPLEXITY OF CELLS. INL DRAWING NUMBER 200-0601-00-098-105687. ALTERNATE ID NUMBER 4289-20-301. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP601) ...

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

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP-601) LOOKING NORTHWEST. PHOTO TAKEN FROM MIDDLE OF CORRIDOR. INL PHOTO NUMBER HD-50-2-3. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP601) ...

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

INTERIOR PHOTO OF MAIN PROCESSING BUILDING PROCESS MAKEUP AREA (CPP-601) LOOKING SOUTH. PHOTO TAKEN FROM MIDDLE OF CORRIDOR. INL PHOTO NUMBER HD-50-3-2. Mike Crane, Photographer, 6/2005 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

10 CFR Appendix D to Subpart D of... - Classes of Actions That Normally Require EISs

Code of Federal Regulations, 2010 CFR

2010-01-01

... average megawatts or more over a 12 month period. This applies to power marketing operations and to siting... Systems D2. Siting/construction/operation/decommissioning of nuclear fuel reprocessing facilities D3. Siting/construction/operation/decommissioning of uranium enrichment facilities D4. Siting/construction...

Experimental studies at the Idaho Chemical Processing Plant on actinide partitioning from acidic nuclear wastes

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

McIssaac, L. D.; Baker, J. D.; Meikrantz, D. H.

1980-01-01

Wastes generated at ICPP and in the reprocessing of LWR fuel is discussed separately. DHDECMP is used as extractant. Studies on DHDECMP purification and toxicity, diluent effects, reaction kinetics, radioloysis, mixer-settler performance, etc. are reported. 10 tables, 3 figures. (DLC)

10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

Code of Federal Regulations, 2012 CFR

2012-01-01

... executing. III. Design Control Measures shall be established to assure that applicable regulatory... control of design interfaces and for coordination among participating design organizations. These measures..., approval, release, distribution, and revision of documents involving design interfaces. The design control...

10 CFR Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

Code of Federal Regulations, 2013 CFR

2013-01-01

... executing. III. Design Control Measures shall be established to assure that applicable regulatory... control of design interfaces and for coordination among participating design organizations. These measures..., approval, release, distribution, and revision of documents involving design interfaces. The design control...

Enhanced electroanalysis in lithium potassium eutectic (LKE) using microfabricated square microelectrodes.

PubMed

Corrigan, Damion K; Blair, Ewen O; Terry, Jonathan G; Walton, Anthony J; Mount, Andrew R

2014-11-18

Molten salts (MSs) are an attractive medium for chemical and electrochemical processing and as a result there is demand for MS-compatible analysis technologies. However, MSs containing redox species present a challenging environment in which to perform analytical measurements because of their corrosive nature, significant thermal convection and the high temperatures involved. This paper outlines the fabrication and characterization of microfabricated square microelectrodes (MSMs) designed for electrochemical analysis in MS systems. Their design enables precise control over electrode dimension, the minimization of stress because of differential thermal expansion through design for high temperature operation, and the minimization of corrosive attack through effective insulation. The exemplar MS system used for characterization was lithium chloride/potassium chloride eutectic (LKE), which has potential applications in pyrochemical nuclear fuel reprocessing, metal refining, molten salt batteries and electric power cells. The observed responses for a range of redox ions between 400 and 500 °C (673 and 773 K) were quantitative and typical of microelectrodes. MSMs also showed the reduced iR drop, steady-state diffusion-limited response, and reduced sensitivity to convection seen for microelectrodes under ambient conditions and expected for these electrodes in comparison to macroelectrodes. Diffusion coefficients were obtained in close agreement with literature values, more readily and at greater precision and accuracy than both macroelectrode and previous microelectrode measurements. The feasibility of extracting individual physical parameters from mixtures of redox species (as required in reprocessing) and of the prolonged measurement required for online monitoring was also demonstrated. Together, this demonstrates that MSMs provide enhanced electrode devices widely applicable to the characterization of redox species in a range of MS systems.

Continental-scale enrichment of atmospheric 14CO2 from the nuclear power industry: potential impact on the estimation of fossil fuel-derived CO2

NASA Astrophysics Data System (ADS)

Graven, H. D.; Gruber, N.

2011-12-01

The 14C-free fossil carbon added to atmospheric CO2 by combustion dilutes the atmospheric 14C/C ratio (Δ14C), potentially providing a means to verify fossil CO2 emissions calculated using economic inventories. However, sources of 14C from nuclear power generation and spent fuel reprocessing can counteract this dilution and may bias 14C/C-based estimates of fossil fuel-derived CO2 if these nuclear influences are not correctly accounted for. Previous studies have examined nuclear influences on local scales, but the potential for continental-scale influences on Δ14C has not yet been explored. We estimate annual 14C emissions from each nuclear site in the world and conduct an Eulerian transport modeling study to investigate the continental-scale, steady-state gradients of Δ14C caused by nuclear activities and fossil fuel combustion. Over large regions of Europe, North America and East Asia, nuclear enrichment may offset at least 20% of the fossil fuel dilution in Δ14C, corresponding to potential biases of more than -0.25 ppm in the CO2 attributed to fossil fuel emissions, larger than the bias from plant and soil respiration in some areas. Model grid cells including high 14C-release reactors or fuel reprocessing sites showed much larger nuclear enrichment, despite the coarse model resolution of 1.8°×1.8°. The recent growth of nuclear 14C emissions increased the potential nuclear bias over 1985-2005, suggesting that changing nuclear activities may complicate the use of Δ14C observations to identify trends in fossil fuel emissions. The magnitude of the potential nuclear bias is largely independent of the choice of reference station in the context of continental-scale Eulerian transport and inversion studies, but could potentially be reduced by an appropriate choice of reference station in the context of local-scale assessments.

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

Casella, Amanda J.; Hylden, Laura R.; Campbell, Emily L.

Knowledge of real-time solution properties and composition is a necessity for any spent nuclear fuel reprocessing method. Metal-ligand speciation in aqueous solutions derived from the dissolved commercial spent fuel is highly dependent upon the acid concentration/pH, which influences extraction efficiency and the resulting speciation in the organic phase. Spectroscopic process monitoring capabilities, incorporated in a counter current centrifugal contactor bank, provide a pathway for on-line real-time measurement of solution pH. The spectroscopic techniques are process-friendly and can be easily configured for on-line applications, while classic potentiometric pH measurements require frequent calibration/maintenance and have poor long-term stability in aggressive chemical andmore » radiation environments. Our research is focused on developing a general method for on-line determination of pH of aqueous solutions through chemometric analysis of Raman spectra. Interpretive quantitative models have been developed and validated under the range of chemical composition and pH using a lactic acid/lactate buffer system. The developed model was applied to spectra obtained on-line during solvent extractions performed in a centrifugal contactor bank. The model predicted the pH within 11% for pH > 2, thus demonstrating that this technique could provide the capability of monitoring pH on-line in applications such as nuclear fuel reprocessing.« less

Sources of the transuranic elements plutonium and neptunium in arctic marine sediments.

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

Cooper, L. W.; Kelley, J. M.; Bond, L. A.

2000-01-01

We report here thermal ionization mass spectrometry measurements of {sup 239}Pu, {sup 240}Pu, {sup 241}Pu, {sup 242}Pu, and {sup 237}Np isolated from oceanic, estuarine, and riverine sediments from the Arctic Ocean Basin. {sup 238}Pu/{sup 239+240}Pu activity ratios are also reported for alpha spectrometric analyses undertaken on a subset of these samples. Our results indicate that the Pu in sediments on the Alaskan shelf and slope, as well as that in the deep basins (Amerasian and Eurasian) of the Arctic Ocean, has its origin in stratospheric and tropospheric fallout. Sediments from the Ob and Yenisei Rivers show isotopic Pu signatures thatmore » are distinctly different from those of northern-hemisphere stratospheric fallout and indicate the presence of weapons-grade Pu originating from nuclear fuel reprocessing wastes generated at Russian facilities within these river catchments. Consequently, sediments of the Eurasian Arctic Ocean, particularly those in the Barents and Kara Seas, probably contain a mixture of Pu from stratospheric fallout, tropospheric fallout, and fuel-reprocessing wastes of riverine origin. In particular, the {sup 241}Pu/{sup 239}Pu ratios observed in these sediments are inconsistent with significant contributions of Pu to the arctic sediments studied from western European reprocessing facilities, principally Sellafield in the UK. Several other potential sources of Pu to arctic sediments can also be excluded as significant based upon the transuranic isotope ratios presented.« less

Characterization of Used Nuclear Fuel with Multivariate Analysis for Process Monitoring

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

Dayman, Kenneth J.; Coble, Jamie B.; Orton, Christopher R.

2014-01-01

The Multi-Isotope Process (MIP) Monitor combines gamma spectroscopy and multivariate analysis to detect anomalies in various process streams in a nuclear fuel reprocessing system. Measured spectra are compared to models of nominal behavior at each measurement location to detect unexpected changes in system behavior. In order to improve the accuracy and specificity of process monitoring, fuel characterization may be used to more accurately train subsequent models in a full analysis scheme. This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict fuel burnup. Nuclide activities for prototypic usedmore » fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this work are artificially low, because the models were trained, optimized, and tested on simulated, noise-free data. However, these results indicate that the developed models may generalize well to new data and that the proposed approach constitutes a viable first step in developing a fuel characterization algorithm based on gamma spectra.« less

A Pebble-Bed Breed-and-Burn Reactor

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

Greenspan, Ehud

2016-03-31

The primary objective of this project is to use three-dimensional fuel shuffling in order to reduce the minimum peak radiation damage of ~550 dpa present Breed-and-Burn (B&B) fast nuclear reactor cores designs (they feature 2-D fuel shuffling) call for to as close as possible to the presently accepted value of 200 dpa thereby enabling earlier commercialization of B&B reactors which could make substantial contribution to energy sustainability and economic stability without need for fuel recycling. Another objective is increasing the average discharge burnup for the same peak discharge burnup thereby (1) increasing the fuel utilization of 2-D shuffled B&B reactorsmore » and (2) reducing the reprocessing capacity required to support a given capacity of FRs that are to recycle fuel.« less

The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

2017-01-01

The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

Redundancy of Supply in the International Nuclear Fuel Fabrication Market: Are Fabrication Services Assured?

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

Seward, Amy M.; Toomey, Christopher; Ford, Benjamin E.

2011-11-14

For several years, Pacific Northwest National Laboratory (PNNL) has been assessing the reliability of nuclear fuel supply in support of the U.S. Department of Energy/National Nuclear Security Administration. Three international low enriched uranium reserves, which are intended back up the existing and well-functioning nuclear fuel market, are currently moving toward implementation. These backup reserves are intended to provide countries credible assurance that of the uninterrupted supply of nuclear fuel to operate their nuclear power reactors in the event that their primary fuel supply is disrupted, whether for political or other reasons. The efficacy of these backup reserves, however, may bemore » constrained without redundant fabrication services. This report presents the findings of a recent PNNL study that simulated outages of varying durations at specific nuclear fuel fabrication plants. The modeling specifically enabled prediction and visualization of the reactors affected and the degree of fuel delivery delay. The results thus provide insight on the extent of vulnerability to nuclear fuel supply disruption at the level of individual fabrication plants, reactors, and countries. The simulation studies demonstrate that, when a reasonable set of qualification criteria are applied, existing fabrication plants are technically qualified to provide backup fabrication services to the majority of the world's power reactors. The report concludes with an assessment of the redundancy of fuel supply in the nuclear fuel market, and a description of potential extra-market mechanisms to enhance the security of fuel supply in cases where it may be warranted. This report is an assessment of the ability of the existing market to respond to supply disruptions that occur for technical reasons. A forthcoming report will address political disruption scenarios.« less

A two-dimensional, finite-difference model of the oxidation of a uranium carbide fuel pellet

NASA Astrophysics Data System (ADS)

Shepherd, James; Fairweather, Michael; Hanson, Bruce C.; Heggs, Peter J.

2015-12-01

The oxidation of spent uranium carbide fuel, a candidate fuel for Generation IV nuclear reactors, is an important process in its potential reprocessing cycle. However, the oxidation of uranium carbide in air is highly exothermic. A model has therefore been developed to predict the temperature rise, as well as other useful information such as reaction completion times, under different reaction conditions in order to help in deriving safe oxidation conditions. Finite difference-methods are used to model the heat and mass transfer processes occurring during the reaction in two dimensions and are coupled to kinetics found in the literature.

Radioactive Waste Management, its Global Implication on Societies, and Political Impact

NASA Astrophysics Data System (ADS)

Matsui, Kazuaki

2009-05-01

Reprocessing plant in Rokkasho, Japan is under commissioning at the end of 2008, and it starts soon to reprocess about 800 Mt of spent fuel per annum, which have been stored at each nuclear power plant sites in Japan. Fission products together with minor actinides separated from uranium and plutonium in the spent fuel contain almost all radioactivity of it and will be vitrified with glass matrix, which then will fill the canisters. The canisters with the high level radioactive waste (HLW) are so hot in both thermal and radiological meanings that they have to be cooled off for decades before bringing out to any destination. Where is the final destination for HLW in Japan, which is located at the rim of the Pacific Ocean with volcanoes? Although geological formation in Japan is not so static and rather active as the other parts of the planet, experts concluded with some intensive studies and researches that there will be a lot of variety of geological formations even in Japan which can host the HLW for so long times of more than million years. Then an organization to implement HLW disposal program was set up and started to campaign for volunteers to accept the survey on geological suitability for HLW disposal. Some local governments wanted to apply, but were crashed down by local and neighbor governments and residents. The above development is not peculiar only to Japan, but generally speaking more or less common for those with radioactive waste programs. This is why the radioactive waste management is not any more science and technology issue but socio-political one. It does not mean further R&D on geological disposal is not any more necessary, but rather we, each of us, should face much more sincerely the societal and political issues caused by the development of the science and technology. Second topic might be how effective partitioning and transformation technology may be to reduce the burden of waste disposal and denature the waste toxicity? The third one might be the proposal of international nuclear fuel centers which supply nuclear fuel to the nuclear power plants in the region and take back spent fuel which will be reprocessed to recover useful energy resources of uranium and plutonium. This may help non proliferation issue due to world nuclear development beyond renaissance.

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

Bruffey, S. H.; Spencer, B. B.; Strachan, D. M.

Four radionuclides have been identified as being sufficiently volatile in the reprocessing of nuclear fuel that their gaseous release needs to be controlled to meet regulatory requirements (Jubin et al. 2011, 2012). These radionuclides are 3H, 14C, 85Kr, and 129I. Of these, 129I has the longest half-life and potentially high biological impact. Accordingly, control of the release of 129I is most critical with respect to the regulations for the release of radioactive material in stack emissions. It is estimated that current EPA regulations (EPA 2010) would require any reprocessing plant in the United States to limit 129I release to lessmore » than 0.05 Ci/MTIHM for a typical fuel burnup of 55 gigawatt days per metric tonne (GWd/t) (Jubin 2011). The study of inorganic iodide in off-gas systems has been almost exclusively limited to I2 and the focus of organic iodide studies has been CH3I. In this document, we provide the results of an examination of publically available literature that is relevant to the presence and sources of both inorganic and organic iodine-bearing species in reprocessing plants. We especially focus on those that have the potential to be poorly sequestered with traditional capture methodologies. Based on the results of the literature survey and some limited thermodynamic modeling, the inorganic iodine species hypoiodous acid (HOI) and iodine monochloride (ICl) were identified as potentially low-sorbing iodine species that could present in off-gas systems. Organic species of interest included both short chain alkyl iodides such as methyl iodide (CH3I) and longer alkyl iodides up to iodododecane (C10H21I). It was found that fuel dissolution may provide conditions conducive to HOI formation and has been shown to result in volatile long-chain alkyl iodides, though these may not volatilize until later in the reprocessing sequence. Solvent extraction processes were found to be significant sources of various organic iodine-bearing species; formation of these was facilitated by the presence of radiolytic decomposition products resulting from radiolysis of tri-n-butyl phosphate and dodecane. Primarily inorganic iodine compounds were expected from waste management processes, including chlorinated species such as ICl. Critical knowledge gaps that must still be addressed include confirmation of the existence and quantification of low-sorbing species in the off-gas of reprocessing facilities. The contributions from penetrating forms of iodine to the plant DF are largely unknown and highly dependent on the magnitude of their presence. These species are likely to be more difficult to remove and it is likely that their sequestration could be improved through the use of different sorbents, through design modifications of the off-gas capture system, or through chemical conversion prior to iodine abatement that would produce more easily captured forms.« less

Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

PubMed

Edwards, Geoffrey W R; Priest, Nicholas D

2014-11-01

The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low as 1 mSv. In addition, if this method is extended so that Pu is also measured, then the combined amount of Pu and Pu is sufficiently high in the thorium-plutonium fuel that a committed effective dose of 1 mSv would be measurable. However, the fraction of Pu and Pu in the other two fuels is sufficiently low that a 1 mSv dose would remain below the detection limit using this technique. Thus new methods, such as fecal measurements of Pu (or other alpha emitters), will be required to measure exposure to these new fuels.

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

Exploratory Development of New and Improved Self-Sealing Materials for Fuel Lines

DTIC Science & Technology

1974-10-01

identify hy block number) New and improved self-sealing fuel line composites were developed under this program. Fabric reinforced plastic and nonflowering...integrated aluminum foil, fabric reinforced laminated fuel line composites employing compressed natural rubber foam as the sealant were fabricated which...successfully sealed wounds inflicted by .30 and .50 caliber projectiles. The weight of these new self-sealing fuel line composites ranged from 0.83

JPRS Report, Proliferation Issues

DTIC Science & Technology

1993-06-07

Ruta Skatikaite; RESPUBLIKA, 19 May 93] ................................................................................ 20 Radioactive Beryllium...nuclear fuel will be transported around 2000 to a reprocessing facility in [Yi] IAEA surveillance cameras are said to take four photos Tokai, Ibaraki...Comparing these two methods, the method of extracting May 93 pp 342-346. plutonium is similar to carrying a backpack to transport goods, while

FLOOR PLAN OF MAIN PROCESSING BUILDING (CPP601) BASEMENT SHOWING PROCESS ...

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

FLOOR PLAN OF MAIN PROCESSING BUILDING (CPP-601) BASEMENT SHOWING PROCESS CORRIDOR AND EIGHTEEN CELLS. TO LEFT IS LABORATORY BUILDING (CPP-602). INL DRAWING NUMBER 200-0601-00-706-051981. ALTERNATE ID NUMBER CPP-E-1981. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP640) LOOKING NORTHEAST ...

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

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP-640) LOOKING NORTHEAST SHOWING DECK FORMING FOR SOUTH SECTION OF OPERATING CORRIDOR; CONSTRUCTION 44 PERCENT COMPLETE. INL PHOTO NUMBER NRTS-60-3624. Holmes, Photographer, 7/25/1960 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

SECOND FLOOR PLAN OF REMOTE ANALYTICAL FACILITY (CPP627) WARM LABORATORY ...

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

SECOND FLOOR PLAN OF REMOTE ANALYTICAL FACILITY (CPP-627) WARM LABORATORY ROOM, DECONTAMINATION ROOM, HOT CHEMISTRY LABORATORY, AND MULTICURIE CELL ROOM. INL DRAWING NUMBER 200-0627-00-098-105066. ALTERNATE ID NUMBER 4272-14-103. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP640) LOOKING NORTHWEST, ...

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

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CPP-640) LOOKING NORTHWEST, SHOWING FORMING FOR NORTH WALLS OF CELLS 1, 4 AND 5; CONSTRUCTION 21 PERCENT COMPLETE. INL PHOTO NUMBER NRTS-60-1874. Holmes, Photographer, 4/21/1960 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

SOUTH SECTION OF WEST ELEVATION OF MAIN PROCESSING BUILDING (CPP601) ...

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

SOUTH SECTION OF WEST ELEVATION OF MAIN PROCESSING BUILDING (CPP-601) LOOKING EAST. HEADEND PLANT BUILDING (CPP-640) APPEARS ON LEFT IN PHOTO. INL PHOTO NUMBER HD-22-3-3. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

OBLIQUE PHOTO OF NORTH AND WEST ELEVATIONS OF REMOTE ANALYTICAL ...

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

OBLIQUE PHOTO OF NORTH AND WEST ELEVATIONS OF REMOTE ANALYTICAL FACILITY (CPP-627) LOOKING SOUTHEAST. LABORATORY AND OFFICE BUILDING (CPP-602) APPEAR ON LEFT IN PHOTO. INL PHOTO NUMBER HD-22-2-2. Mike Crane, Photographer, 11/1998 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

76 FR 40943 - Notice of Issuance of Regulatory Guide

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-12

..., Revision 3, ``Criteria for Use of Computers in Safety Systems of Nuclear Power Plants.'' FOR FURTHER..., ``Criteria for Use of Computers in Safety Systems of Nuclear Power Plants,'' was issued with a temporary... Fuel Reprocessing Plants,'' to 10 CFR part 50 with regard to the use of computers in safety systems of...

Radiation chemistry for modern nuclear energy development

NASA Astrophysics Data System (ADS)

Chmielewski, Andrzej G.; Szołucha, Monika M.

2016-07-01

Radiation chemistry plays a significant role in modern nuclear energy development. Pioneering research in nuclear science, for example the development of generation IV nuclear reactors, cannot be pursued without chemical solutions. Present issues related to light water reactors concern radiolysis of water in the primary circuit; long-term storage of spent nuclear fuel; radiation effects on cables and wire insulation, and on ion exchangers used for water purification; as well as the procedures of radioactive waste reprocessing and storage. Radiation effects on materials and enhanced corrosion are crucial in current (II/III/III+) and future (IV) generation reactors, and in waste management, deep geological disposal and spent fuel reprocessing. The new generation of reactors (III+ and IV) impose new challenges for radiation chemists due to their new conditions of operation and the usage of new types of coolant. In the case of the supercritical water-cooled reactor (SCWR), water chemistry control may be the key factor in preventing corrosion of reactor structural materials. This paper mainly focuses on radiation effects on long-term performance and safety in the development of nuclear power plants.

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), long term care, reduced access to 'dirty' bomb materials, the social and political costs of siting new facilities and the psychological impact of no solution to the nuclear waste problem, were taken into account, the costs would be far lower than those of the present fuel cycle. (authors)« less

US-RERTR Advanced Fuel Development Plans : 1999.

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

Meyer, M. K.

1998-10-22

Twelve fuel alloys were included in the very-high-density RERTR-1 and RERTR-2 microplate irradiation experiments. Experience gained during fabrication and results from the post-irradiation examination of these fuels has allowed us to narrow the focus of our fuel development efforts in preparation for the next set of irradiation experiments. Specific technical problems in both the areas of fuel fabrication and irradiation performance remain to be addressed. Examples of these are powder fabrication, fuel phase gamma stability versus density, and fuel-matrix interaction. In order to more efficiently address metal alloy fuel performance issues, work will continue on establishing a theoretical basis formore » alloy stability and metal alloy dispersion fuel irradiation performance. Plans to address these fuel development issues in the coming year will be presented.« less

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

DOE PAGES

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; ...

2016-07-15

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo havemore » similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21« less

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

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

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo havemore » similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21« less

Development of Online Spectroscopic pH Monitoring for Nuclear Fuel Reprocessing Plants: Weak Acid Schemes.

PubMed

Casella, Amanda J; Ahlers, Laura R H; Campbell, Emily L; Levitskaia, Tatiana G; Peterson, James M; Smith, Frances N; Bryan, Samuel A

2015-05-19

In nuclear fuel reprocessing, separating trivalent minor actinides and lanthanide fission products is extremely challenging and often necessitates tight pH control in TALSPEAK (Trivalent Actinide-Lanthanide Separation by Phosphorus reagent Extraction from Aqueous Komplexes) separations. In TALSPEAK and similar advanced processes, aqueous pH is one of the most important factors governing the partitioning of lanthanides and actinides between an aqueous phase containing a polyaminopolycarboxylate complexing agent and a weak carboxylic acid buffer and an organic phase containing an acidic organophosphorus extractant. Real-time pH monitoring would significantly increase confidence in the separation performance. Our research is focused on developing a general method for online determination of the pH of aqueous solutions through chemometric analysis of Raman spectra. Spectroscopic process-monitoring capabilities, incorporated in a counter-current centrifugal contactor bank, provide a pathway for online, real-time measurement of solution pH. The spectroscopic techniques are process-friendly and can be easily configured for online applications, whereas classic potentiometric pH measurements require frequent calibration/maintenance and have poor long-term stability in aggressive chemical and radiation environments. Raman spectroscopy discriminates between the protonated and deprotonated forms of the carboxylic acid buffer, and the chemometric processing of the Raman spectral data with PLS (partial least-squares) regression provides a means to quantify their respective abundances and therefore determine the solution pH. Interpretive quantitative models have been developed and validated under a range of chemical composition and pH conditions using a lactic acid/lactate buffer system. The developed model was applied to new spectra obtained from online spectral measurements during a solvent extraction experiment using a counter-current centrifugal contactor bank. The model predicted the pH of this validation data set within 11% for pH > 2, thus demonstrating that this technique could provide the capability of monitoring pH online in applications such as nuclear fuel reprocessing.

Yttrium and rare earth stabilized fast reactor metal fuel

DOEpatents

Guon, Jerold; Grantham, LeRoy F.; Specht, Eugene R.

1992-01-01

To increase the operating temperature of a reactor, the melting point and mechanical properties of the fuel must be increased. For an actinide-rich fuel, yttrium, lanthanum and/or rare earth elements can be added, as stabilizers, to uranium and plutonium and/or a mixture of other actinides to raise the melting point of the fuel and improve its mechanical properties. Since only about 1% of the actinide fuel may be yttrium, lanthanum, or a rare earth element, the neutron penalty is low, the reactor core size can be reduced, the fuel can be burned efficiently, reprocessing requirements are reduced, and the nuclear waste disposal volumes reduced. A further advantage occurs when yttrium, lanthanum, and/or other rare earth elements are exposed to radiation in a reactor, they produce only short half life radioisotopes, which reduce nuclear waste disposal problems through much shorter assured-isolation requirements.

Viscosity Meaurement Technique for Metal Fuels

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

Ban, Heng; Kennedy, Rory

2015-02-09

Metallic fuels have exceptional transient behavior, excellent thermal conductivity, and a more straightforward reprocessing path, which does not separate out pure plutonium from the process stream. Fabrication of fuel containing minor actinides and rare earth (RE) elements for irradiation tests, for instance, U-20Pu-3Am-2Np-1.0RE-15Zr samples at the Idaho National Laboratory, is generally done by melt casting in an inert atmosphere. For the design of a casting system and further scale up development, computational modeling of the casting process is needed to provide information on melt flow and solidification for process optimization. Therefore, there is a need for melt viscosity data, themore » most important melt property that controls the melt flow. The goal of the project was to develop a measurement technique that uses fully sealed melt sample with no Americium vapor loss to determine the viscosity of metallic melts and at temperatures relevant to the casting process. The specific objectives of the project were to: develop mathematical models to establish the principle of the measurement method, design and build a viscosity measurement prototype system based on the established principle, and calibrate the system and quantify the uncertainty range. The result of the project indicates that the oscillation cup technique is applicable for melt viscosity measurement. Detailed mathematical models of innovative sample ampoule designs were developed to not only determine melt viscosity, but also melt density under certain designs. Measurement uncertainties were analyzed and quantified. The result of this project can be used as the initial step toward the eventual goal of establishing a viscosity measurement system for radioactive melts.« less

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

Shepherd, James; Fairweather, Michael; Hanson, Bruce C.

The oxidation of spent uranium carbide fuel, a candidate fuel for Generation IV nuclear reactors, is an important process in its potential reprocessing cycle. However, the oxidation of uranium carbide in air is highly exothermic. A model has therefore been developed to predict the temperature rise, as well as other useful information such as reaction completion times, under different reaction conditions in order to help in deriving safe oxidation conditions. Finite difference-methods are used to model the heat and mass transfer processes occurring during the reaction in two dimensions and are coupled to kinetics found in the literature.

Analysis of the factors that impact the reliability of high level waste canister materials

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

Boyd, W.K.; Hall, A.M.

1977-09-19

The analysis encompassed identification and analysis of potential threats to canister integrity arising in the course of waste solidification, interim storage at the fuels reprocessing plant, wet and dry shipment, and geologic storage. Fabrication techniques and quality assurance requirements necessary to insure optimum canister reliability were considered taking into account such factors as welding procedure, surface preparation, stress relief, remote weld closure, and inspection methods. Alternative canister materials and canister systems were also considered in terms of optimum reliability in the face of threats to the canister's integrity, ease of fabrication, inspection, handling and cost. If interim storage in airmore » is admissible, the sequence suggested comprises producing a glass-type waste product in a continuous ceramic melter, pouring into a carbon steel or low-alloy steel canister of moderately heavy wall thickness, storing in air upright on a pad and surrounded by a concrete radiation shield, and thereafter placing in geologic storage without overpacking. Should the decision be to store in water during the interim period, then use of either a 304 L stainless steel canister overpacked with a solution-annealed and fast-cooled 304 L container, or a single high-alloy canister, is suggested. The high alloy may be Inconel 600, Incoloy Alloy 800, or Incoloy Alloy 825. In either case, it is suggested that the container be overpacked with a moderately heavy wall carbon steel or low-alloy steel cask for geologic storage to ensure ready retrievability. 19 figs., 5 tables.« less

Radionuclide speciation in effluent from La Hague reprocessing plant in France.

PubMed

Salbu, B; Skipperud, L; Germain, P; Guéguéniat, P; Strand, P; Lind, O C; Christensen, G

2003-09-01

Effluent from the La Hague nuclear fuel reprocessing plant was mixed with seawater in order to investigate the fate of the various radionuclides. Thus, a major objective of the present work is to characterize the effluent from La Hague reprocessing plant and to study how the radionuclide speciation changes with time when discharged into the marine environment. Discharges from the La Hague nuclear reprocessing plant represent an important source of artificially produced radionuclides to the North Sea. The transport, distribution, and biological uptake of radionuclides in the marine environment depends, however, on the physicochemical forms of radionuclides in the discharged effluents and on transformation processes that occur after entering the coastal waters. Information of these processes is needed to understand the transport and long-term distribution of the radionuclides. In the present work, a weekly discharged effluent from the nuclear fuel reprocessing plant at Cap La Hague in France was mixed with coastal water and fractionated with respect to particle size and charged species using ultra centrifugation and hollow fiber ultrafiltration with on line ion exchange. The size distribution pattern of gamma-emitting radionuclides was followed during a 62-h period after mixing the effluent with seawater. 54Mn was present as particulate material in the effluent, while other investigated radionuclides were discharged in a more mobile form or were mobilized after mixing with sea water (e.g., 60Co) and can be transported long distances in the sea. Sediments can act as a sink for less mobile discharged radionuclides (Skipperud et al. 2000). A kinetic model experiment was performed to provide information of the time-dependent distribution coefficients, Kd (t). The retention of the effluent radionuclides in sediments was surprisingly low (Kd 20-50), and the sediments acted as a poor sink for the released radionuclides. Due to the presence of non-reacting radionuclide species in the effluent, a major fraction of the radionuclides, such as Cs-isotopes, 106Ru and 125Sb, in the effluent will be subjected to marine transport to the Northern Seas (i.e., the North Sea, Norwegian Sea and the Barents Sea). The La Hague effluent may, therefore, contribute to enriched levels of radionuclides found in the English Channel, including 90Sr, 60Co and Pu-isotopes, and also 106Ru and 125Sb.

Significance of and prospects for fuel recycle in Japan

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

Otsuka, K.; Ikeda, K.

Japan's nuclear power plant capacity ranks fourth in the world at around 20 GW. But nuclear fuel cycle industries (enrichment, reprocessing and radioactive waste management) are still in their infancy compared with the size and stage of the power plants. Thus it is a matter of urgency to establish a nuclear fuel cycle in Japan which can promote nuclear energy as a quasi-indigenous energy source. Some moves toward establishing a nuclear fuel cycle have been observed recently. As a case in point, in July 1984, the Federation of Electric Power Companies has formally requested Aomori Prefecture to locate nuclear fuelmore » cycle facilities in the Shimokita Peninsula region. Plutonium recovered from spent fuel will be utilized in LWR, ATR, and FBR. Research and development activities on these technologies are in progress.« less

Fabrication of thorium bearing carbide fuels

DOEpatents

Gutierrez, R.L.; Herbst, R.J.; Johnson, K.W.R.

Thorium-uranium carbide and thorium-plutonium carbide fuel pellets have been fabricated by the carbothermic reduction process. Temperatures of 1750/sup 0/C and 2000/sup 0/C were used during the reduction cycle. Sintering temperatures of 1800/sup 0/C and 2000/sup 0/C were used to prepare fuel pellet densities of 87% and > 94% of theoretical, respectively. The process allows the fabrication of kilogram quantities of fuel with good reproductibility of chemical and phase composition.

Mixed Oxide Fresh Fuel Package Auxiliary Equipment

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

Yapuncich, F.; Ross, A.; Clark, R.H.

2008-07-01

The United States Department of Energy's National Nuclear Security Administration (NNSA) is overseeing the construction the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF) on the Savannah River Site. The new facility, being constructed by NNSA's contractor Shaw AREVA MOX Services, will fabricate fuel assemblies utilizing surplus plutonium as feedstock. The fuel will be used in designated commercial nuclear reactors. The MOX Fresh Fuel Package (MFFP), which has recently been licensed by the Nuclear Regulatory Commission (NRC) as a type B package (USA/9295/B(U)F-96), will be utilized to transport the fabricated fuel assemblies from the MFFF to the nuclear reactors. It wasmore » necessary to develop auxiliary equipment that would be able to efficiently handle the high precision fuel assemblies. Also, the physical constraints of the MFFF and the nuclear power plants require that the equipment be capable of loading and unloading the fuel assemblies both vertically and horizontally. The ability to reconfigure the load/unload evolution builds in a large degree of flexibility for the MFFP for the handling of many types of both fuel and non fuel payloads. The design and analysis met various technical specifications including dynamic and static seismic criteria. The fabrication was completed by three major fabrication facilities within the United States. The testing was conducted by Sandia National Laboratories. The unique design specifications and successful testing sequences will be discussed. (authors)« less

Observations Derived From the Characterization of Monolithic Fuel Plates Irradiated as Part of the RERTR-6 Experiment

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

D. D. Keiser, Jr.; A. B. Robinson; M. R. Finlay

2007-09-01

Evaluation of the PIE results of the monolithic plates that were irradiated as part of the RERTR-6 experiment has continued. Specifically, comparisons have been made between the microstructures of the fuel plates before and after irradiation. Using the results from the rigorous characterization that was performed on the as-fabricated plates using scanning electron microscopy, it is possible to improve understanding of how monolithic fuel plates perform when they are irradiated. This paper will discuss the changes that occur, if any, in the microstructure of a monolithic fuel plate that is fabricated using techniques like what were employed for fabricating RERTR-6more » fuel plates. In addition, the performance of fuel/cladding interaction layers that were present in the fuel plates due to the fabrication process will be discussed, particularly in the context of swelling of these layers and how these layers exhibit different behaviors depending on whether the fuel alloy in the fuel plate is U-7Mo or U-10Mo.« less

A proliferation of nuclear waste for the Southeast.

PubMed

Alvarez, Robert; Smith, Stephen

2007-12-01

The U.S. Department of Energy's (DOE) Global Nuclear Energy Partnership (GNEP) is being promoted as a program to bring about the expansion of worldwide nuclear energy. Here in the U.S. much of this proposed nuclear power expansion is slated to happen in the Southeast, including here in South Carolina. Under the GNEP plan, the United States and its nuclear partners would sell nuclear power plants to developing nations that agree not to pursue technologies that would aid nuclear weapons production, notably reprocessing and uranium enrichment. As part of the deal, the United States would take highly radioactive spent ("used") fuel rods to a reprocessing center in this country. Upon analysis of the proposal, it is clear that DOE lacks a credible plan for the safe management and disposal of radioactive wastes stemming from the GNEP program and that the high costs and possible public health and environmental impacts from the program pose significant risks, especially to this region. Given past failures to address waste problems before they were created, DOE's rush to invest major public funds for deployment of reprocessing should be suspended.

Canister arrangement for storing radioactive waste

DOEpatents

Lorenzo, D.K.; Van Cleve, J.E. Jr.

1980-04-23

The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

Canister arrangement for storing radioactive waste

DOEpatents

Lorenzo, Donald K.; Van Cleve, Jr., John E.

1982-01-01

The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

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

R. Wigeland; T. Taiwo; M. Todosow

The recently completed comprehensive evaluation and screening of nuclear fuel cycle options identified a number of potentially promising fuel cycles for R&D that offer what could be considered by decision-makers as having the potential for significant improvement compared to the current U.S. fuel cycle. The fuel cycles that consistently performed the best were recycle fuel cycles that used self-sustaining fast reactors operating with either U/Pu or U/TRU recycle fuel and also included options where the fast reactors provided fissile materials to support operation of thermal reactors. However, based on the evaluation criteria and metrics used in the study, there wasmore » no difference in benefit between recycle of U/Pu and U/TRU (where TRU is plutonium and the minor actinides) while there were differences in the challenges for developing and deploying such fuel cycles, with U/TRU recycle being more challenging. This observation prompted the question as to the desirability of pursuing R&D on U/TRU recycle given that there may not be an increase in benefit. As a result, activities have been pursued to further investigate the performance differences between U/Pu and U/TRU recycle based on considering issues beyond those used in the evaluation and screening study to identify, if possible, areas where there are significant benefits of U/TRU recycle compared to U/Pu recycle. These new considerations focused on several areas, but especially on the impact on disposal of the HLW, which in the case of U/Pu recycle contains all of the minor actinides along with fission products, while in the case of U/TRU recycle only contains the losses of minor actinides from the reprocessing and recycle fuel fabrication operations. This difference in content has several implications. One impact is on the time dependent decay heat which can affect handling and the use of space in a geologic repository. Another impact concerns the HLW form and volume, since presence of minor actinides may adversely affect the ability to reduce HLW volume. The short-term radioactivity and long-term radiotoxicity of the HLW is also affected, which may be of more or less importance depending on the specific geologic disposal environment. To study these potential effects, a range of waste forms and disposal environments were used in the analysis, documenting to what extent the recycle of minor actinides in addition to plutonium may offer further benefit. Another area of investigation concerned the recycle fuel, for the fast reactor and for the thermal reactors they may support. Information to date indicates that U/Pu fuel may be simpler to fabricate and has a much more extensive database than U/TRU fuel, one of the reasons for the increased challenge for developing and deploying a U/TRU fuel cycle, and also indicates that heterogeneous recycle of the minor actinides may be even more difficult as compared to homogeneous recycle. This information was reviewed and updated to reflect the most recent studies for the purpose of informing on all aspects of the differences between U/Pu and U/TRU recycle. The results of all of these investigations will be presented to provide information on the findings concerning the value of U/TRU recycle.« less

Fabrication and testing of U-7Mo monolithic plate fuel with Zircaloy cladding

NASA Astrophysics Data System (ADS)

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; Wachs, D. M.; Finlay, M. R.

2016-10-01

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U-(7-10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry-4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry-4 clad U-7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry-4 and U-(7-10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction-either from fabrication or in-reactor testing-and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak).

Fabric-based alkaline direct formate microfluidic fuel cells.

PubMed

Domalaon, Kryls; Tang, Catherine; Mendez, Alex; Bernal, Franky; Purohit, Krutarth; Pham, Linda; Haan, John; Gomez, Frank A

2017-04-01

Fabric-based microfluidic fuel cells (MFCs) serve as a novel, cost-efficient alternative to traditional FCs and batteries, since fluids naturally travel across fabric via capillary action, eliminating the need for an external pump and lowering production and operation costs. Building on previous research with Y-shaped paper-based MFCs, fabric-based MFCs mitigate fragility and durability issues caused by long periods of fuel immersion. In this study, we describe a microfluidic fabric-based direct formate fuel cell, with 5 M potassium formate and 30% hydrogen peroxide as the anode fuel and cathode oxidant, respectively. Using a two-strip, stacked design, the optimized parameters include the type of encasement, the barrier, and the fabric type. Surface contact of the fabric and laminate sheet expedited flow and respective chemical reactions. The maximum current (22.83 mA/cm 2 ) and power (4.40 mW/cm 2 ) densities achieved with a 65% cotton/35% polyester blend material are a respective 8.7% and 32% higher than previous studies with Y-shaped paper-based MFCs. In series configuration, the MFCs generate sufficient energy to power a handheld calculator, a thermometer, and a spectrum of light-emitting diodes. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Demand driven salt clean-up in a molten salt fast reactor - Defining a priority list.

PubMed

Merk, B; Litskevich, D; Gregg, R; Mount, A R

2018-01-01

The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified.

Bonded polyimide fuel cell package

DOEpatents

Morse, Jeffrey D.; Jankowski, Alan; Graff, Robert T.; Bettencourt, Kerry

2010-06-08

Described herein are processes for fabricating microfluidic fuel cell systems with embedded components in which micron-scale features are formed by bonding layers of DuPont Kapton.TM. polyimide laminate. A microfluidic fuel cell system fabricated using this process is also described.

Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

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

Lippek, H.E.; Schuller, C.R.

1979-03-01

A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light mostmore » of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel.« less

Inert matrix fuel in dispersion type fuel elements

NASA Astrophysics Data System (ADS)

Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

2006-06-01

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

2017-12-01

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

Fabrication of capsule assemblies, phase 3

NASA Technical Reports Server (NTRS)

Keeton, A. R.; Stemann, L. G.

1973-01-01

Thirteen capsule assemblies were fabricated for evaluation of fuel pin design concepts for a fast spectrum lithium cooled compact space power reactor. These instrumented assemblies were designed for real time test of prototype fuel pins. Uranium mononitride fuel pins were encased in AISI 304L stainless steel capsules. Fabrication procedures were fully qualified by process development and assembly qualification tests. Instrumentation reliability was achieved utilizing specially processed and closely controlled thermocouple hot zone fabrication and by thermal screening tests. Overall capsule reliability was achieved with an all electron beam welded assembly.

Composite nuclear fuel fabrication methodology for gas fast reactors

NASA Astrophysics Data System (ADS)

Vasudevamurthy, Gokul

An advanced fuel form for use in Gas Fast Reactors (GFR) was investigated. Criteria for the fuel includes operation at high temperature (˜1400°C) and high burnup (˜150 MWD/MTHM) with effective retention of fission products even during transient temperatures exceeding 1600°C. The GFR fuel is expected to contain up to 20% transuranics for a closed fuel cycle. Earlier evaluations of reference fuels for the GFR have included ceramic-ceramic (cercer) dispersion type composite fuels of mixed carbide or nitride microspheres coated with SiC in a SiC matrix. Studies have indicated that ZrC is a potential replacement for SiC on account of its higher melting point, increased fission product corrosion resistance and better chemical stability. The present work investigated natural uranium carbide microspheres in a ZrC matrix instead of SiC. Known issues of minor actinide volatility during traditional fabrication procedures necessitated the investigation of still high temperature but more rapid fabrication techniques to minimize these anticipated losses. In this regard, fabrication of ZrC matrix by combustion synthesis from zirconium and graphite powders was studied. Criteria were established to obtain sufficient matrix density with UC microsphere volume fractions up to 30%. Tests involving production of microspheres by spark erosion method (similar to electrodischarge machining) showed the inability of the method to produce UC microspheres in the desired range of 300 to 1200 mum. A rotating electrode device was developed using a minimum current of 80A and rotating at speeds up to 1500 rpm to fabricate microspheres between 355 and 1200 mum. Using the ZrC process knowledge, UC electrodes were fabricated and studied for use in the rotating electrode device to produce UC microspheres. Fabrication of the cercer composite form was studied using microsphere volume fractions of 10%, 20%, and 30%. The macrostructure of the composite and individual components at various stages were characterized to understand the required fabrication techniques and at the same time meet the necessary GFR fuel characteristics.

Bonded polyimide fuel cell package and method thereof

DOEpatents

Morse, Jeffrey D.; Jankowski, Alan; Graff, Robert T.; Bettencourt, Kerry

2005-11-01

Described herein are processes for fabricating microfluidic fuel cell systems with embedded components in which micron-scale features are formed by bonding layers of DuPont Kapton.TM. polyimide laminate. A microfluidic fuel cell system fabricated using this process is also described.

Method of preparation of bonded polyimide fuel cell package

DOEpatents

Morse, Jeffrey D [Martinez, CA; Jankowski, Alan [Livermore, CA; Graff, Robert T [Modesto, CA; Bettencourt, Kerry [Dublin, CA

2011-04-26

Described herein are processes for fabricating microfluidic fuel cell systems with embedded components in which micron-scale features are formed by bonding layers of DuPont Kapton.TM. polyimide laminate. A microfluidic fuel cell system fabricated using this process is also described.

Nuclear physics research operation. Monthly report, November 1958

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

Faulkner, J.E.

1958-12-10

This report is a summary of projects worked on in support of the production reactors at Hanford. The projects include criticality studies, from tasks associated with fuel element reprocessing to shipments of slightly enriched uranium. They include studies of neutron cross sections for different reactions and neutron flux measurements in different reactor locations, as well as design studies for future reactor projects.

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Code of Federal Regulations, 2010 CFR

2010-01-01

... transferred to a Federal repository no later than 10 years following separation of fission products from the.... Disposal of high-level radioactive fission product waste material will not be permitted on any land other... of the policy stated above with respect to high-level radioactive fission product wastes generated...

10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

Code of Federal Regulations, 2011 CFR

2011-01-01

... transferred to a Federal repository no later than 10 years following separation of fission products from the.... Disposal of high-level radioactive fission product waste material will not be permitted on any land other... of the policy stated above with respect to high-level radioactive fission product wastes generated...

No increased cancer risks from nuclear facilities

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

Not Available

1990-11-08

This article reports the results of a US Department of Health and Human Services (HHS) and National Cancer Institute (NCI) two-year survey that shows no increased risk of death from cancer for people living in counties containing or close to nuclear plants. 62 plants and their surrounding counties were included in the survey including commercial, US DOE and fuel reprocessing plants.

Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

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

M. D. Staiger

2007-06-01

This report provides a quantitative inventory and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. From December 1963 through May 2000, liquid radioactive wastes generated by spent nuclear fuel reprocessing were converted into a solid, granular form called calcine. This report also contains a description of the calcine storage bins.

Liquidus temperature and chemical durability of selected glasses to immobilize rare earth oxides waste

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

Mohd Fadzil, Syazwani Binti; Hrma, Pavel R.; Schweiger, Michael J.

Pyroprocessing is a reprocessing method for managing and reusing used nuclear fuel (UNF) by dissolving it in an electrorefiner with a molten alkali or alkaline earth chloride salt mixture while avoiding wet reprocessing. Pyroprocessing UNF with a LiCl-KCl eutectic salt releases the fission products from the fuel and generates a variety of metallic and salt-based species, including rare earth (RE) chlorides. If the RE-chlorides are converted to oxides, borosilicate glass is a prime candidate for their immobilization because of its durability and ability to dissolve almost any RE waste component into the matrix at high loadings. Crystallization that occurs inmore » waste glasses as the waste loading increases may complicate glass processing and affect the product quality. This work compares three types of borosilicate glasses in terms of liquidus temperature (TL): the International Simple Glass designed by the International Working Group, sodium borosilicate glass developed by Korea Hydro and Nuclear Power, and the lanthanide aluminoborosilicate (LABS) glass established in the United States. The LABS glass allows the highest waste loadings (over 50 mass% RE2O3) while possessing an acceptable chemical durability.« less

Corrosion study of a highly durable electrolyzer based on cold crucible technique for pyrochemical reprocessing of spent nuclear oxide fuel

NASA Astrophysics Data System (ADS)

Takeuchi, M.; Arai, Y.; Kase, T.; Nakajima, Y.

2013-01-01

The application of the cold crucible technique to a pyrochemical electrolyzer used in the oxide-electrowinning method, which is a method for the pyrochemical reprocessing of spent nuclear oxide fuel, is proposed as a means for improving corrosion resistance. The electrolyzer suffers from a severe corrosion environment consisting of molten salt and corrosive gas. In this study, corrosion tests for several metals in molten 2CsCl-NaCl at 923 K with purging chlorine gas were conducted under controlled material temperature conditions. The results revealed that the corrosion rates of several materials were significantly decreased by the material cooling effect. In particular, Hastelloy C-22 showed excellent corrosion resistance with a corrosion rate of just under 0.01 mm/y in both molten salt and vapor phases by controlling the material surface at 473 K. Finally, an engineering-scale crucible composed of Hastelloy C-22 was manufactured to demonstrate the basic function of the cold crucible. The cold crucible induction melting system with the new concept Hastelloy crucible showed good compatibility with respect to its heating and cooling performances.

Plutonium: Advancing our Understanding to Support Sustainable Nuclear Fuel Cycles

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

Lines, Amanda M.; Adami, Susan R.; Casella, Amanda

With Global energy needs increasing, real energy solutions to meet demands now, are needed. Fossil fuels are not an ideal candidate to meet these needs because of their negative impact on the environment. Renewables such as wind and solar have huge potential, but still need major technological advancements (particularly in the area of battery storage) before they can effectively meet growing world needs. The best option for meeting large energy needs without a large carbon footprint is nuclear energy. Of course, nuclear energy can face a fair amount of opposition and concern. However, through modern engineering and science many ofmore » these concerns can now be addressed. Many safety concerns can be met by engineering advancements, but perhaps the biggest area of concern is what to do with the used nuclear fuel after it is removed from the reactor. Currently the United States (and several other countries) utilize an open fuel cycle, meaning fuel is only used once and then discarded. It should be noted that fuel coming out of a reactor has utilized approximately 1% of the total energy that could be produced by the uranium in the fuel rod. The answer here is to close the fuel cycle and recycle the nuclear materials. By reprocessing used nuclear fuel, all the U can be repurposed without requiring disposal. The various fission products can be removed and either discarded (hugely reduced waste volume) or more reasonably, utilized in specialty reactors to make more energy or needed research/medical isotopes. While reprocessing technology is currently advanced enough to meet energy needs, completing research to improve and better understand these techniques is still needed. Better understanding behavior of fission products is one area of important research. Despite it being discovered over 75 years ago, plutonium is still an exciting element to study because of the complex solution chemistry it exhibits. In aqueous solutions Pu can exist simultaneously in multiple oxidation states, including 3+, 4+, and 6+. It also readily forms a variety of metal-ligand complexes depending on solution pH and available ligands. Understanding of the behavior of Pu in solution remains an important area of research today, with relevance to developing sustainable nuclear fuel cycles, minimizing its impact on the environment, and detecting and preventing the spread of nuclear weapons technology.« less

76 FR 65544 - Standard Format and Content of License Applications for Mixed Oxide Fuel Fabrication Facilities

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-21

...The U.S. Nuclear Regulatory Commission (NRC or Commission) is issuing a revision to regulatory guide (RG) 3.39, ``Standard Format and Content of License Applications for Mixed Oxide Fuel Fabrication Facilities.'' This guide endorses the standard format and content for license applications and integrated safety analysis (ISA) summaries described in the current version of NUREG-1718, ``Standard Review Plan for the Review of an Application for a Mixed Oxide (MOX) Fuel Fabrication Facility,'' as a method that the NRC staff finds acceptable for meeting the regulatory requirements of Title 10 of the Code of Federal Regulations (10 CFR) part 70, ``Domestic Licensing of Special Nuclear Material'' for mixed oxide fuel fabrication facilities.

Simulation of ground-water flow near the nuclear-fuel reprocessing facility at the Western New York Nuclear Service Center, Cattaraugus County, New York

USGS Publications Warehouse

Yager, R.M.

1987-01-01

A two-dimensional finite-difference model was developed to simulate groundwater flow in a surficial sand and gravel deposit underlying the nuclear fuel reprocessing facility at Western New York Nuclear Service Center near West Valley, N.Y. The sand and gravel deposit overlies a till plateau that abuts an upland area of siltstone and shale on its west side, and is bounded on the other three sides by deeply incised stream channels that drain to Buttermilk Creek, a tributary to Cattaraugus Creek. Radioactive materials are stored within the reprocessing plant and are also buried within a till deposit at the facility. Tritiated water is stored in a lagoon system near the plant and released under permit to Franks Creek, a tributary to Buttermilk Creek. Groundwater levels predicted by steady-state simulations closely matched those measured in 23 observation wells, with an average error of 0.5 meter. Simulated groundwater discharges to two stream channels and a subsurface drain were within 5% of recorded values. Steady-state simulations used an average annual recharge rate of 46 cm/yr; predicted evapotranspiration loss from the ground was 20 cm/yr. The lateral range in hydraulic conductivity obtained through model calibration was 0.6 to 10 m/day. Model simulations indicated that 33% of the groundwater discharged from the sand and gravel unit (2.6 L/sec) is lost by evapotranspiration, 3% (3.0 L/sec) flows to seepage faces at the periphery of the plateau, 20% (1.6 L/sec) discharges to stream channels that drain a large wetland area near the center of the plateau, and the remaining 8% (0.6 L/sec) discharges to a subsurface french drain and to a wastewater treatment system. Groundwater levels computed by a transient-state simulation of an annual climatic cycle, including seasonal variation in recharge and evapotranspiration, closely matched water levels measured in eight observation wells. The model predicted that the subsurface drain and the stream channel that drains the wetland would intercept most of the recharge originating near the reprocessing plant. (Lantz-PTT)

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

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.

Determination of the NPP Kr\\vsko spent fuel decay heat

NASA Astrophysics Data System (ADS)

Kromar, Marjan; Kurinčič, Bojan

2017-07-01

Nuclear fuel is designed to support fission process in a reactor core. Some of the isotopes, formed during the fission, decay and produce decay heat and radiation. Accurate knowledge of the nuclide inventory producing decay heat is important after reactor shut down, during the fuel storage and subsequent reprocessing or disposal. In this paper possibility to calculate the fuel isotopic composition and determination of the fuel decay heat with the Serpent code is investigated. Serpent is a well-known Monte Carlo code used primarily for the calculation of the neutron transport in a reactor. It has been validated for the burn-up calculations. In the calculation of the fuel decay heat different set of isotopes is important than in the neutron transport case. Comparison with the Origen code is performed to verify that the Serpent is taking into account all isotopes important to assess the fuel decay heat. After the code validation, a sensitivity study is carried out. Influence of several factors such as enrichment, fuel temperature, moderator temperature (density), soluble boron concentration, average power, burnable absorbers, and burnup is analyzed.

Solid polymer electrolyte (SPE) fuel cell technology program, phase 1/1A. [design and fabrication

NASA Technical Reports Server (NTRS)

1975-01-01

A solid polymer electrolyte fuel cell was studied for the purpose of improving the characteristics of the technology. Several facets were evaluated, namely: (1) reduced fuel cell costs; (2) reduced fuel cell weight; (3) improved fuel cell efficiency; and (4) increased systems compatibility. Demonstrated advances were incorporated into a full scale hardware design. A single cell unit was fabricated. A substantial degree of success was demonstrated.

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

Luther, Erik Paul; Leckie, Rafael M.; Dombrowski, David E.

This supplemental report describes fuel fabrication efforts conducted for the Idaho National Laboratory Trade Study for the TREAT Conversion project that is exploring the replacement of the HEU (Highly Enriched Uranium) fuel core of the TREAT reactor with LEU (Low Enriched Uranium) fuel. Previous reports have documented fabrication of fuel by the “upgrade” process developed at Los Alamos National Laboratory. These experiments supplement an earlier report that describes efforts to increase the graphite content of extruded fuel and minimize cracking.

Experience with soluble neutron poisons for criticality control at ICPP

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

Wilson, R.E.; Mortimer, S.R.

1978-01-01

Soluble neutron poisons assure criticality control in two of the headend fuel reprocessing systems at the Idaho Chemical Processing Plant. Soluble poisons have been used successfully since 1964 and will be employed in the projected new headend processes. The use of soluble poisons (1) greatly increases the process output (2) allows versatility in the size of fuel assemblies processed and (3) allows the practical reprocessing of some fuels. The safety limit for all fluids entering the U-Zr alloy dissolver is 3.6 g/liter boron. To allow for possible deviations in the measurement systems and drift between analytical sampling periods, the standardmore » practice is to use 3.85 g/liter boron as the lower limit. This dissolver has had 4000 successful hours of operation using soluble poisons. The electrolytic dissolution process depends on soluble gadolinium for criticality safety. This system is used to process high enriched uranium clad in stainless steel. Electrolytic dissolution takes advantage of the anodic corrosion that occurs when a large electrical current is passed through the fuel elements in a corrosive environment. Three control methods are used on each headend system. First, the poison is mixed according to standard operating procedures and the measurements are affirmed by the operator's supervisor. Second, the poisoned solution is stirred, sampled, analyzed, and the analysis reported while still in the mix tank. Finally, a Nuclear Poison Detection System (NPDS) must show an acceptable poison concentration before the solution can be transferred. The major disadvantage of using soluble poisons is the need for very sophisticated control systems and procedures, which require extensive checkout. The need for a poisoned primary heating and cooling system means a secondary system is needed as well. Experience has shown, however, that production enhancement more than makes up for the problems.« less

Demand driven salt clean-up in a molten salt fast reactor – Defining a priority list

PubMed Central

Litskevich, D.; Gregg, R.; Mount, A. R.

2018-01-01

The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified. PMID:29494604

The processing of enriched germanium for the Majorana Demonstrator and R&D for a next generation double-beta decay experiment

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

Abgrall, N.; Arnquist, I. J.; Avignone III, F. T.

The Majorana Demonstrator is an array of point-contact Ge detectors fabricated from Ge isotopically enriched to 88% in 76Ge to search for neutrinoless double beta decay. The processing of Ge for germanium detectors is a well-known technology. However, because of the high cost of Ge enriched in 76Ge special procedures were required to maximize the yield of detector mass and to minimize exposure to cosmic rays. These procedures include careful accounting for the material; shielding it to reduce cosmogenic generation of radioactive isotopes; and development of special reprocessing techniques for contaminated solid germanium, shavings, grindings, acid etchant and cutting fluidsmore » from detector fabrication. Processing procedures were developed that resulted in a total yield in detector mass of 70%. However, none of the acid-etch solution and only 50% of the cutting fluids from detector fabrication were reprocessed. Had they been processed, the projections for the recovery yield would be between 80% and 85%. Maximizing yield is critical to justify a possible future ton-scale experiment. A process for recovery of germanium from the acid-etch solution was developed with yield of about 90%. All material was shielded or stored underground whenever possible to minimize the formation of 68Ge by cosmic rays, which contributes background in the double-beta decay region of interest and cannot be removed by zone refinement and crystal growth. Formation of 68Ge was reduced by a significant factor over that in natural abundance detectors not protected from cosmic rays.« less

The processing of enriched germanium for the Majorana   Demonstrator  and R&D for a next generation double-beta decay experiment

DOE PAGES

Abgrall, N.; Arnquist, I. J.; Avignone III, F. T.; ...

2017-10-07

The Majorana Demonstrator is an array of point-contact Ge detectors fabricated from Ge isotopically enriched to 88% in 76Ge to search for neutrinoless double beta decay. The processing of Ge for germanium detectors is a well-known technology. However, because of the high cost of Ge enriched in 76Ge special procedures were required to maximize the yield of detector mass and to minimize exposure to cosmic rays. These procedures include careful accounting for the material; shielding it to reduce cosmogenic generation of radioactive isotopes; and development of special reprocessing techniques for contaminated solid germanium, shavings, grindings, acid etchant and cutting fluidsmore » from detector fabrication. Processing procedures were developed that resulted in a total yield in detector mass of 70%. However, none of the acid-etch solution and only 50% of the cutting fluids from detector fabrication were reprocessed. Had they been processed, the projections for the recovery yield would be between 80% and 85%. Maximizing yield is critical to justify a possible future ton-scale experiment. A process for recovery of germanium from the acid-etch solution was developed with yield of about 90%. All material was shielded or stored underground whenever possible to minimize the formation of 68Ge by cosmic rays, which contributes background in the double-beta decay region of interest and cannot be removed by zone refinement and crystal growth. Formation of 68Ge was reduced by a significant factor over that in natural abundance detectors not protected from cosmic rays.« less

The processing of enriched germanium for the MAJORANA DEMONSTRATOR and R&D for a next generation double-beta decay experiment

NASA Astrophysics Data System (ADS)

Abgrall, N.; Arnquist, I. J.; Avignone, F. T., III; Barabash, A. S.; Bertrand, F. E.; Bradley, A. W.; Brudanin, V.; Busch, M.; Buuck, M.; Caja, J.; Caja, M.; Caldwell, T. S.; Christofferson, C. D.; Chu, P.-H.; Cuesta, C.; Detwiler, J. A.; Dunagan, C.; Dunstan, D. T.; Efremenko, Yu.; Ejiri, H.; Elliott, S. R.; Gilliss, T.; Giovanetti, G. K.; Goett, J.; Green, M. P.; Gruszko, J.; Guinn, I. S.; Guiseppe, V. E.; Haufe, C. R. S.; Henning, R.; Hoppe, E. W.; Jasinski, B. R.; Kidd, M. F.; Konovalov, S. I.; Kouzes, R. T.; Lopez, A. M.; MacMullin, J.; Martin, R. D.; Massarczyk, R.; Meijer, S. J.; Mertens, S.; Meyer, J. H.; Myslik, J.; O'Shaughnessy, C.; Poon, A. W. P.; Radford, D. C.; Rager, J.; Reine, A. L.; Reising, J. A.; Rielage, K.; Robertson, R. G. H.; Shanks, B.; Shirchenko, M.; Suriano, A. M.; Tedeschi, D.; Toth, L. M.; Trimble, J. E.; Varner, R. L.; Vasilyev, S.; Vetter, K.; Vorren, K.; White, B. R.; Wilkerson, J. F.; Wiseman, C.; Xu, W.; Yakushev, E.; Yu, C.-H.; Yumatov, V.; Zhitnikov, I.; Zhu, B. X.

2018-01-01

The MAJORANA DEMONSTRATOR is an array of point-contact Ge detectors fabricated from Ge isotopically enriched to 88% in 76 Ge to search for neutrinoless double beta decay. The processing of Ge for germanium detectors is a well-known technology. However, because of the high cost of Ge enriched in 76 Ge special procedures were required to maximize the yield of detector mass and to minimize exposure to cosmic rays. These procedures include careful accounting for the material; shielding it to reduce cosmogenic generation of radioactive isotopes; and development of special reprocessing techniques for contaminated solid germanium, shavings, grindings, acid etchant and cutting fluids from detector fabrication. Processing procedures were developed that resulted in a total yield in detector mass of 70%. However, none of the acid-etch solution and only 50% of the cutting fluids from detector fabrication were reprocessed. Had they been processed, the projections for the recovery yield would be between 80% and 85%. Maximizing yield is critical to justify a possible future ton-scale experiment. A process for recovery of germanium from the acid-etch solution was developed with yield of about 90%. All material was shielded or stored underground whenever possible to minimize the formation of 68Ge by cosmic rays, which contributes background in the double-beta decay region of interest and cannot be removed by zone refinement and crystal growth. Formation of 68Ge was reduced by a significant factor over that in natural abundance detectors not protected from cosmic rays.

The processing of enriched germanium for the Majorana Demonstrator and R&D for a next generation double-beta decay experiment

DOE PAGES

Abgrall, N.; Arnquist, I. J.; Avignone III, F. T.; ...

2017-10-07

The Majorana Demonstrator is an array of point-contact Ge detectors fabricated from Ge isotopically enriched to 88% in 76Ge to search for neutrinoless double beta decay. The processing of Ge for germanium detectors is a well-known technology. However, because of the high cost of Ge enriched in 76Ge special procedures were required to maximize the yield of detector mass and to minimize exposure to cosmic rays. These procedures include careful accounting for the material; shielding it to reduce cosmogenic generation of radioactive isotopes; and development of special reprocessing techniques for contaminated solid germanium, shavings, grindings, acid etchant and cutting fluidsmore » from detector fabrication. Processing procedures were developed that resulted in a total yield in detector mass of 70%. However, none of the acid-etch solution and only 50% of the cutting fluids from detector fabrication were reprocessed. Had they been processed, the projections for the recovery yield would be between 80% and 85%. Maximizing yield is critical to justify a possible future ton-scale experiment. A process for recovery of germanium from the acid-etch solution was developed with yield of about 90%. All material was shielded or stored underground whenever possible to minimize the formation of 68Ge by cosmic rays, which contributes background in the double-beta decay region of interest and cannot be removed by zone refinement and crystal growth. Formation of 68Ge was reduced by a significant factor over that in natural abundance detectors not protected from cosmic rays.« less

Reprocessing and Recycling of Highly Cross-Linked Ion-Conducting Networks through Transalkylation Exchanges of C-N Bonds.

PubMed

Obadia, Mona M; Mudraboyina, Bhanu P; Serghei, Anatoli; Montarnal, Damien; Drockenmuller, Eric

2015-05-13

Exploiting exchangeable covalent bonds as dynamic cross-links recently afforded a new class of polymer materials coined as vitrimers. These permanent networks are insoluble and infusible, but the network topology can be reshuffled at high temperatures, thus enabling glasslike plastic deformation and reprocessing without depolymerization. We disclose herein the development of functional and high-value ion-conducting vitrimers that take inspiration from poly(ionic liquid)s. Tunable networks with high ionic content are obtained by the solvent- and catalyst-free polyaddition of an α-azide-ω-alkyne monomer and simultaneous alkylation of the resulting poly(1,2,3-triazole)s with a series of difunctional cross-linking agents. Temperature-induced transalkylation exchanges of C-N bonds between 1,2,3-triazolium cross-links and halide-functionalized dangling chains enable recycling and reprocessing of these highly cross-linked permanent networks. They can also be recycled by depolymerization with specific solvents able to displace the transalkylation equilibrium, and they display a great potential for applications that require solid electrolytes with excellent mechanical performances and facile processing such as supercapacitors, batteries, fuel cells, and separation membranes.

NORTH SECTION OF WEST ELEVATION OF MAIN PROCESSING BUILDING (CPP601) ...

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

NORTH SECTION OF WEST ELEVATION OF MAIN PROCESSING BUILDING (CPP-601) LOOKING EAST. HOT PILOT PLANT BUILDING (CPP-640) APPEARS IN RIGHT OF PHOTO. THE REMOTE ANALYTICAL FACILITY (CPP-627) WAS LOCATED ON CONCRETE PAD IN FOREGROUND. INL PHOTO NUMBER HD-54-33-3. Mike Crane, Photographer, 7/2006 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

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

Muhammad Abir; Fahima Islam; Hyoung Koo Lee

2014-11-01

The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the Highmore » Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.« less

THE ATTRACTIVENESS OF MATERIALS IN ADVANCED NUCLEAR FUEL CYCLES FOR VARIOUS PROLIFERATION AND THEFT SCENARIOS

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

Bathke, C. G.; Ebbinghaus, Bartley B.; Collins, Brian A.

2012-08-29

We must anticipate that the day is approaching when details of nuclear weapons design and fabrication will become common knowledge. On that day we must be particularly certain that all special nuclear materials (SNM) are adequately accounted for and protected and that we have a clear understanding of the utility of nuclear materials to potential adversaries. To this end, this paper examines the attractiveness of materials mixtures containing SNM and alternate nuclear materials associated with the plutonium-uranium reduction extraction (Purex), uranium extraction (UREX), coextraction (COEX), thorium extraction (THOREX), and PYROX (an electrochemical refining method) reprocessing schemes. This paper provides amore » set of figures of merit for evaluating material attractiveness that covers a broad range of proliferant state and subnational group capabilities. The primary conclusion of this paper is that all fissile material must be rigorously safeguarded to detect diversion by a state and must be provided the highest levels of physical protection to prevent theft by subnational groups; no 'silver bullet' fuel cycle has been found that will permit the relaxation of current international safeguards or national physical security protection levels. The work reported herein has been performed at the request of the U.S. Department of Energy (DOE) and is based on the calculation of 'attractiveness levels' that are expressed in terms consistent with, but normally reserved for, the nuclear materials in DOE nuclear facilities. The methodology and findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security is discussed.« less

Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

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

Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their high-enriched uranium spentmore » nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.« less

Fabrication of Monolithic RERTR Fuels by Hot Isostatic Pressing

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

Jan-Fong Jue; Blair H. Park; Curtis R. Clark

2010-11-01

The RERTR (Reduced Enrichment for Research and Test Reactors) Program is developing advanced nuclear fuels for high-power test reactors. Monolithic fuel design provides higher uranium loading than that of the traditional dispersion fuel design. Hot isostatic pressing is a promising process for low-cost batch fabrication of monolithic RERTR fuel plates for these high-power reactors. Bonding U Mo fuel foil and 6061 Al cladding by hot isostatic press bonding was successfully developed at Idaho National Laboratory. Due to the relatively high processing temperature, the interaction between fuel meat and aluminum cladding is a concern. Two different methods were employed to mitigatemore » this effect: (1) a diffusion barrier and (2) a doping addition to the interface. Both types of fuel plates have been fabricated by hot isostatic press bonding. Preliminary results show that the direct fuel/cladding interaction during the bonding process was eliminated by introducing a thin zirconium diffusion barrier layer between the fuel and the cladding. Fuel plates were also produced and characterized with a silicon-rich interlayer between fuel and cladding. This paper reports the recent progress of this developmental effort and identifies the areas that need further attention.« less

Casting evaluation of U-Zr alloy system fuel slug for SFR prepared by injection casting method

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

Song, Hoon; Kim, Jong-Hwan; Kim, Ki-Hwan

2013-07-01

Metal fuel slugs of U-Pu-Zr alloys for Sodium-cooled Fast Reactor (SFR) have conventionally been fabricated by a vacuum injection casting method. Recently, management of minor actinides (MA) became an important issue because direct disposal of the long-lived MA can be a long-term burden for a tentative repository up to several hundreds of thousand years. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long-lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. In order tomore » prevent the evaporation of volatile elements such as Am, alternative fabrication methods of metal fuel slugs have been studied applying gravity casting, and improved injection casting in KAERI, including melting under inert atmosphere. And then, metal fuel slugs were examined with casting soundness, density, chemical analysis, particle size distribution and microstructural characteristics. Based on these results there is a high level of confidence that Am losses will also be effectively controlled by application of a modest amount of overpressure. A surrogate fuel slug was generally soundly cast by improved injection casting method, melted fuel material under inert atmosphere.« less

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

Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.

Here, this paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li 2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl 2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersivemore » X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability.« less

Impact of minor actinide recycling on sustainable fuel cycle options

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

Heidet, F.; Kim, T. K.; Taiwo, T. A.

The recent Evaluation and Screening study chartered by the U.S. Department of Energy, Office of Nuclear Energy, has identified four fuel cycle options as being the most promising. Among these four options, the two single-stage fuel cycles rely on a fast reactor and are differing in the fact that in one case only uranium and plutonium are recycled while in the other case minor actinides are also recycled. The two other fuel cycles are two-stage and rely on both fast and thermal reactors. They also differ in the fact that in one case only uranium and plutonium are recycled whilemore » in the other case minor actinides are also recycled. The current study assesses the impact of recycling minor actinides on the reactor core design, its performance characteristics, and the characteristics of the recycled material and waste material. The recycling of minor actinides is found not to affect the reactor core performance, as long as the same cycle length, core layout and specific power are being used. One notable difference is that the required transuranics (TRU) content is slightly increased when minor actinides are recycled. The mass flows are mostly unchanged given a same specific power and cycle length. Although the material mass flows and reactor performance characteristics are hardly affected by recycling minor actinides, some differences are observed in the waste characteristics between the two fuel cycles considered. The absence of minor actinides in the waste results in a different buildup of decay products, and in somewhat different behaviors depending on the characteristic and time frame considered. Recycling of minor actinides is found to result in a reduction of the waste characteristics ranging from 10% to 90%. These results are consistent with previous studies in this domain and depending on the time frame considered, packaging conditions, repository site, repository strategy, the differences observed in the waste characteristics could be beneficial and help improve the repository performance. On the other hand, recycling minor actinides also results in an increase of the recycled fuel characteristics and therefore of the charged fuel. The radioactivity is slightly increased while the decay heat and radiotoxicities are very significantly increased. Despite these differences, the characteristics of the fuel at time of discharge remain similar whether minor actinides are recycled or not, with the exception of the inhalation radiotoxicity which is significantly larger with minor actinide recycling. After some cooling the characteristics of the discharged fuel become larger when minor actinides are recycled, potentially affecting the reprocessing plant requirements. Recycling minor actinides has a negative impact on the characteristics of the fresh fuel and will make it more challenging to fabricate fuel containing minor actinides.« less

3S (Safeguards, Security, Safety) based pyroprocessing facility safety evaluation plan

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

Ku, J.H.; Choung, W.M.; You, G.S.

The big advantage of pyroprocessing for the management of spent fuels against the conventional reprocessing technologies lies in its proliferation resistance since the pure plutonium cannot be separated from the spent fuel. The extracted materials can be directly used as metal fuel in a fast reactor, and pyroprocessing reduces drastically the volume and heat load of the spent fuel. KAERI has implemented the SBD (Safeguards-By-Design) concept in nuclear fuel cycle facilities. The goal of SBD is to integrate international safeguards into the entire facility design process since the very beginning of the design phase. This paper presents a safety evaluationmore » plan using a conceptual design of a reference pyroprocessing facility, in which 3S (Safeguards, Security, Safety)-By-Design (3SBD) concept is integrated from early conceptual design phase. The purpose of this paper is to establish an advanced pyroprocessing hot cell facility design concept based on 3SBD for the successful realization of pyroprocessing technology with enhanced safety and proliferation resistance.« less

Fabrication of simulated DUPIC fuel

NASA Astrophysics Data System (ADS)

Kang, Kweon Ho; Song, Ki Chan; Park, Hee Sung; Moon, Je Sun; Yang, Myung Seung

2000-12-01

Simulated DUPIC fuel provides a convenient way to investigate the DUPIC fuel properties and behavior such as thermal conductivity, thermal expansion, fission gas release, leaching, and so on without the complications of handling radioactive materials. Several pellets simulating the composition and microstructure of DUPIC fuel are fabricated by resintering the powder, which was treated through OREOX process of simulated spent PWR fuel pellets, which had been prepared from a mixture of UO2 and stable forms of constituent nuclides. The key issues for producing simulated pellets that replicate the phases and microstructure of irradiated fuel are to achieve a submicrometre dispersion during mixing and diffusional homogeneity during sintering. This study describes the powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using the simulated spent PWR fuel. The homogeneity of additives in the powder was observed after attrition milling. The microstructure of the simulated spent PWR fuel agrees well with the other studies. The leading structural features observed are as follows: rare earth and other oxides dissolved in the UO2 matrix, small metallic precipitates distributed throughout the matrix, and a perovskite phase finely dispersed on grain boundaries.

10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

Code of Federal Regulations, 2011 CFR

2011-01-01

... performance and safety during reactor operation. Also, in all cases precise control of processes, procedures... performance. (a) Items that are considered especially designed or prepared for the fabrication of fuel... pellets; (2) Automatic welding machines especially designed or prepared for welding end caps onto the fuel...

Effects of the foil flatness on the stress-strain characteristics of U10Mo alloy based monolithic mini-plates

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

Hakan Ozaltun; Pavel Medvedev

The effects of the foil flatness on stress-strain behavior of monolithic fuel mini-plates during fabrication and irradiation were studied. Monolithic plate-type fuels are a new fuel form being developed for research and test reactors to achieve higher uranium densities. This concept facilitates the use of low-enriched uranium fuel in the reactor. These fuel elements are comprised of a high density, low enrichment, U–Mo alloy based fuel foil encapsulated in a cladding material made of Aluminum. To evaluate the effects of the foil flatness on the stress-strain behavior of the plates during fabrication, irradiation and shutdown stages, a representative plate frommore » RERTR-12 experiments (Plate L1P756) was considered. Both fabrication and irradiation processes of the plate were simulated by using actual irradiation parameters. The simulations were repeated for various foil curvatures to observe the effects of the foil flatness on the peak stress and strain magnitudes of the fuel elements. Results of fabrication simulations revealed that the flatness of the foil does not have a considerable impact on the post fabrication stress-strain fields. Furthermore, the irradiation simulations indicated that any post-fabrication stresses in the foil would be relieved relatively fast in the reactor. While, the perfectly flat foil provided the slightly better mechanical performance, overall difference between the flat-foil case and curved-foil case was not significant. Even though the peak stresses are less affected, the foil curvature has several implications on the strain magnitudes in the cladding. It was observed that with an increasing foil curvature, there is a slight increase in the cladding strains.« less

Uranium nitride fuel fabrication for SP-100 reactors

NASA Technical Reports Server (NTRS)

Mason, Richard E.; Chidester, Kenneth M.; Hoth, Carl W.; Matthews, Bruce R.

1987-01-01

Fuel pins of uranium mononitride clad in Nb-1 percent Zr were fabricated for irradiation tests in EBR-II. Laboratory scale process parameters to synthesize UN powders and fabricate UN pellets were developed. Uranium mononitride was prepared by converting UO2 to UN. Fuel pellets were prepared by communition of UN briquettes, uniaxial pressing, and high temperature sintering. Techniques for machining, cleaning, and welding Nb-1 percent Zr cladding components were developed. End caps were electron beam welded to the tubing. Helium back-fill holes were sealed with a laser weld.

Uranium nitride fuel fabrication for SP-100 reactors

NASA Astrophysics Data System (ADS)

Mason, Richard E.; Chidester, Kenneth M.; Hoth, Carl W.; Matthews, Bruce R.

Fuel pins of uranium mononitride clad in Nb-1 percent Zr were fabricated for irradiation tests in EBR-II. Laboratory scale process parameters to synthesize UN powders and fabricate UN pellets were developed. Uranium mononitride was prepared by converting UO2 to UN. Fuel pellets were prepared by communition of UN briquettes, uniaxial pressing, and high temperature sintering. Techniques for machining, cleaning, and welding Nb-1 percent Zr cladding components were developed. End caps were electron beam welded to the tubing. Helium back-fill holes were sealed with a laser weld.

Conversion of radioactive ferrocyanide compounds to immobile glasses

DOEpatents

Schulz, Wallace W.; Dressen, A. Louise

1977-04-26

Complex radioactive ferrocyanide compounds result from the scavenging of cesium from waste products produced in the chemical reprocessing of nuclear fuel. These ferrocyanides, in accordance with this process, are converted to an immobile glass, resistant to leaching by water, by fusion together with sodium carbonate and a mixture of (a) basalt and boron trioxide (B.sub.2 O.sub.3) or (b) silica (SiO.sub.2) and lime (CaO).

International Partnerships to Combat Weapons of Mass Destruction

DTIC Science & Technology

2008-05-01

public, plant , and animal), medicine , economics, law enforcement, and the scientific process itself—and thus engage a range of expert and stakeholder...OCCASIONAL PAPER 6 and replace them with fossil fuel plants ; a multilateral effort to dispose of 34 tons of Russian plutonium; and a number of...nonproliferation benefits: there is no technology ‘silver bullet’ that can be built into an enrichment plant or reprocessing plant that can prevent a

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CP640) LOOKING NORTHWEST ...

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

CONSTRUCTION PROGRESS PHOTO OF HOT PILOT PLANT (CP-640) LOOKING NORTHWEST SHOWING FORMING AND PLACEMENT OF REINFORCING STEEL FOR SOUTH WALLS OF CELLS 1, 3, 4 AND 5 AND WEST WALL FOR CELLS 1 AND 2; CONSTRUCTION 13 PERCENT COMPLETE. INL PHOTO NUMBER NRTS 59-6436. J. Anderson, Photographer, 12/18/1959 - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

TRANSURANIC STUDIES STATUS AND PROBLEM STATEMENT

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

Leuze, R E

1959-04-29

The purpose of the Transuranics Program is to develop separation processes for the transuranic elements, primarily those produced by long-term neutron irradiation of Pu/sup 239/. The program includes laboratory process development, pilot-plant process testing, processing of 10 kg of Pu/sup 239/ irradiated to greater than 99% burn-up for plutonium and americium-curium recovery, and processing the reirradiated plutonium and americium-curium fractions. The proposed method for processing highly irradiated plutonium is: (1) plutonium-aluminum alloy dissolution in HNO/sub 3/; (2) plutonium recovery by TBP extraction; (3) americium, curium, and rare-earth extraction by TBP from neutral nitrate solution; (4) partial rare-earth removal (primarily lanthanum)more » by americium-curium extraction into 100% TBP from 15M HNO/sub 3/; (5) additional rare-earth removal by extraction in 0.48M mono-2-ethylhexylphosphoric acid from 12M HCl; and (6) americium-curium purification by chloride anion exchange. Processing through the 100% TBP, 15M HNO/sub 3/ cycle can be carried out in the Power Reactor Fuel Reprocessing Pilot Plant. New facilities are proposed 15M HNO/ sub 3/ cycle can be carried out in the Power Reactor Fuel Reprocessing Pilot Plant. New facilities are proposed for laboratory process development studies and the final processing of the transplutonic elements. (auth)« less

X-ray Absorption Spectroscopy Investigation of Iodine Capture by Silver-Exchanged Mordenite

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

Abney, Carter W.; Nan, Yue; Tavlarides, Lawrence L.

Capture of radioactive iodine is a significant consideration during reprocessing of spent nuclear fuel and disposal of legacy wastes. While silver-exchanged mordenite (AgZ) is widely regarded as a benchmark material for assessing iodine adsorption performance, previous research efforts have largely focused on bulk material properties rather than the underpinning molecular interactions that achieve effective iodine capture. As a result, the fundamental understanding necessary to identify and mitigate deactivation pathways for the recycle of AgZ is not available. In this paper, we applied X-ray Absorption Fine Structure (XAFS) spectroscopy to investigate AgZ following activation, adsorption of iodine, regeneration, and recycle, observingmore » no appreciable degradation in performance due to the highly controlled conditions under which the AgZ was maintained. Fits of the extended XAFS (EXAFS) data reveal complete formation of Ag 0 nanoparticles upon treatment with H 2, and confirm the formation of α-AgI within the mordenite channels in addition to surface γ/β-AgI nanoparticles following iodine exposure. Analysis of the nanoparticle size and fractional composition of α-AgI to γ/β-AgI supports ripening of surface nanoparticles as a function of recycle. Finally, this work provides a foundation for future investigation of AgZ deactivation under conditions relevant to spent nuclear fuel reprocessing.« less

X-ray Absorption Spectroscopy Investigation of Iodine Capture by Silver-Exchanged Mordenite

DOE PAGES

Abney, Carter W.; Nan, Yue; Tavlarides, Lawrence L.

2017-03-29

Capture of radioactive iodine is a significant consideration during reprocessing of spent nuclear fuel and disposal of legacy wastes. While silver-exchanged mordenite (AgZ) is widely regarded as a benchmark material for assessing iodine adsorption performance, previous research efforts have largely focused on bulk material properties rather than the underpinning molecular interactions that achieve effective iodine capture. As a result, the fundamental understanding necessary to identify and mitigate deactivation pathways for the recycle of AgZ is not available. In this paper, we applied X-ray Absorption Fine Structure (XAFS) spectroscopy to investigate AgZ following activation, adsorption of iodine, regeneration, and recycle, observingmore » no appreciable degradation in performance due to the highly controlled conditions under which the AgZ was maintained. Fits of the extended XAFS (EXAFS) data reveal complete formation of Ag 0 nanoparticles upon treatment with H 2, and confirm the formation of α-AgI within the mordenite channels in addition to surface γ/β-AgI nanoparticles following iodine exposure. Analysis of the nanoparticle size and fractional composition of α-AgI to γ/β-AgI supports ripening of surface nanoparticles as a function of recycle. Finally, this work provides a foundation for future investigation of AgZ deactivation under conditions relevant to spent nuclear fuel reprocessing.« less

129I in the environment of the La Hague nuclear fuel reprocessing plant--from sea to land.

PubMed

Fréchou, C; Calmet, D

2003-01-01

In recent years, particular attention was paid to the long-lived radionuclides discharged with authorized low-level radioactive liquid and gaseous effluents by the nuclear spent fuel reprocessing plants of La Hague and Sellafield. The knowledge of (129)I (half-life=15.7 x 10(6) a) distribution in the environment is required to assess the radiological impact to the environment and population living in the area under the direct influence of La Hague NRP discharges. Measurement difficulties of (129)I in environmental matrices, where it is usually present at trace level, limited data published on (129)I activity levels in the European and more particularly in the French territory. Studies conducted to qualify a new alternative measurement method, direct gamma-X spectrometry with experimental self-absorption correction, led to test samples collected in the La Hague marine and terrestrial environment : seaweeds, lichens, grass, bovine thyroids, etc. All these results, often already published separately for analytical purposes and treated for intercomparison exercises, are presented here together in a radioecological manner. The levels of (129)I activity and (129)I/(127)I ratios in these samples show the spatial and temporal influence of the La Hague NRP in its local near-field environment as well as at the regional scale along the French Channel coast.

FURTHER ASSESSMENTS OF THE ATTRACTIVENESS OF MATERIALS IN ADVANCED NUCLEAR FUEL CYCLES FROM A SAFEGUARDS PERSPECTIVE

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

Bathke, C. G.; Jarvinen, G. D.; Wallace, R. K.

2008-10-01

This paper summarizes the results of an extension to an earlier study [ ] that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with the PUREX, UREX+, and COEX reprocessing schemes. This study focuses on the materials associated with the UREX, COEX, THOREX, and PYROX reprocessing schemes. This study also examines what is required to render plutonium as “unattractive.” Furthermore, combining the results of this study with those from the earlier study permits a comparison of the uranium and thorium based fuel cycles on the basis of the attractiveness of the SNM associated with each fuelmore » cycle. Both studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of “attractiveness levels” that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities [ ]. The methodology and key findings will be presented. Additionally, how these attractiveness levels relate to proliferation resistance (e.g. by increasing impediments to the diversion, theft, undeclared production of SNM for the purpose of acquiring a nuclear weapon), and how they could be used to help inform policy makers, will be discussed.« less

Fabrication of thorium bearing carbide fuels

DOEpatents

Gutierrez, Rueben L.; Herbst, Richard J.; Johnson, Karl W. R.

1981-01-01

Thorium-uranium carbide and thorium-plutonium carbide fuel pellets have been fabricated by the carbothermic reduction process. Temperatures of 1750.degree. C. and 2000.degree. C. were used during the reduction cycle. Sintering temperatures of 1800.degree. C. and 2000.degree. C. were used to prepare fuel pellet densities of 87% and >94% of theoretical, respectively. The process allows the fabrication of kilogram quantities of fuel with good reproducibility of chemicals and phase composition. Methods employing liquid techniques that form carbide microspheres or alloying-techniques which form alloys of thorium-uranium or thorium-plutonium suffer from limitation on the quantities processed of because of criticality concerns and lack of precise control of process conditions, respectively.

Fabrication and Characterization of Surrogate Fuel Particles Using the Spark Erosion Method

NASA Astrophysics Data System (ADS)

Metzger, Kathryn E.

In light of the disaster at the Fukushima Daiichi Nuclear Plant, the Department of Energy's Advanced Fuels Program has shifted its interest from enhanced performance fuels to enhanced accident tolerance fuels. Dispersion fuels possess higher thermal conductivities than traditional light water reactor fuel and as a result, offer improved safety margins. The benefits of a dispersion fuel are due to the presence of the secondary non-fissile phase (matrix), which serves as a barrier to fission products and improves the overall thermal performance of the fuel. However, the presence of a matrix material reduces the fuel volume, which lowers the fissile content of dispersion. This issue can be remedied through the development of higher density fuel phases or through an optimization of fuel particle size and volume loading. The latter requirement necessitates the development of fabrication methods to produce small, micron-order fuel particles. This research examines the capabilities of the spark erosion process to fabricate particles on the order of 10 μm. A custom-built spark erosion device by CT Electromechanica was used to produce stainless steel surrogate fuel particles in a deionized water dielectric. Three arc intensities were evaluated to determine the effect on particle size. Particles were filtered from the dielectric using a polycarbonate membrane filter and vacuum filtration system. Fabricated particles were characterized via field emission scanning electron microscopy (FESEM), laser light particle size analysis, energy-dispersive spectroscopy (EDS), X-ray diffraction analysis (XRD), and gas pycnometry. FESEM images reveal that the spark erosion process produces highly spherical particles on the order of 10 microns. These findings are substantiated by the results of particle size analysis. Additionally, EDS and XRD results indicate the presence of oxide phases, which suggests the dielectric reacted with the molten debris during particle formation.

Year One Summary of X-energy Pebble Fuel Development at ORNL

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

Helmreich, Grant W.; Hunn, John D.; McMurray, Jake W.

2017-06-01

The Advanced Reactor Concepts X-energy (ARC-Xe) Pebble Fuel Development project at Oak Ridge National Laboratory (ORNL) has successfully completed its first year, having made excellent progress in accomplishing programmatic objectives. The primary focus of research at ORNL in support of X-energy has been the training of X-energy fuel fabrication engineers and the establishment of US pebble fuel production capabilities able to supply the Xe-100 pebble-bed reactor. These efforts have been strongly supported by particle fuel fabrication and characterization expertise present at ORNL from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program.

Potential of metal-organic frameworks for separation of xenon and krypton.

PubMed

Banerjee, Debasis; Cairns, Amy J; Liu, Jian; Motkuri, Radha K; Nune, Satish K; Fernandez, Carlos A; Krishna, Rajamani; Strachan, Denis M; Thallapally, Praveen K

2015-02-17

CONSPECTUS: The total world energy demand is predicted to rise significantly over the next few decades, primarily driven by the continuous growth of the developing world. With rapid depletion of nonrenewable traditional fossil fuels, which currently account for almost 86% of the worldwide energy output, the search for viable alternative energy resources is becoming more important from a national security and economic development standpoint. Nuclear energy, an emission-free, high-energy-density source produced by means of controlled nuclear fission, is often considered as a clean, affordable alternative to fossil fuel. However, the successful installation of an efficient and economically viable industrial-scale process to properly sequester and mitigate the nuclear-fission-related, highly radioactive waste (e.g., used nuclear fuel (UNF)) is a prerequisite for any further development of nuclear energy in the near future. Reprocessing of UNF is often considered to be a logical way to minimize the volume of high-level radioactive waste, though the generation of volatile radionuclides during reprocessing raises a significant engineering challenge for its successful implementation. The volatile radionuclides include but are not limited to noble gases (predominately isotopes of Xe and Kr) and must be captured during the process to avoid being released into the environment. Currently, energy-intensive cryogenic distillation is the primary means to capture and separate radioactive noble gas isotopes during UNF reprocessing. A similar cryogenic process is implemented during commercial production of noble gases though removal from air. In light of their high commercial values, particularly in lighting and medical industries, and associated high production costs, alternate approaches for Xe/Kr capture and storage are of contemporary research interest. The proposed pathways for Xe/Kr removal and capture can essentially be divided in two categories: selective absorption by dissolution in solvents and physisorption on porous materials. Physisorption-based separation and adsorption on highly functional porous materials are promising alternatives to the energy-intensive cryogenic distillation process, where the adsorbents are characterized by high surface areas and thus high removal capacities and often can be chemically fine-tuned to enhance the adsorbate-adsorbent interactions for optimum selectivity. Several traditional porous adsorbents such as zeolites and activated carbon have been tested for noble gas capture but have shown low capacity, selectivity, and lack of modularity. Metal-organic frameworks (MOFs) or porous coordination polymers (PCPs) are an emerging class of solid-state adsorbents that can be tailor-made for applications ranging from gas adsorption and separation to catalysis and sensing. Herein we give a concise summary of the background and development of Xe/Kr separation technologies with a focus on UNF reprocessing and the prospects of MOF-based adsorbents for that particular application.

Processing and fabrication of mixed uranium/refractory metal carbide fuels with liquid-phase sintering

NASA Astrophysics Data System (ADS)

Knight, Travis W.; Anghaie, Samim

2002-11-01

Optimization of powder processing techniques were sought for the fabrication of single-phase, solid-solution mixed uranium/refractory metal carbide nuclear fuels - namely (U, Zr, Nb)C. These advanced, ultra-high temperature nuclear fuels have great potential for improved performance over graphite matrix, dispersed fuels tested in the Rover/NERVA program of the 1960s and early 1970s. Hypostoichiometric fuel samples with carbon-to-metal ratios of 0.98, uranium metal mole fractions of 5% and 10%, and porosities less than 5% were fabricated. These qualities should provide for the longest life and highest performance capability for these fuels. Study and optimization of processing methods were necessary to provide the quality assurance of samples for meaningful testing and assessment of performance for nuclear thermal propulsion applications. The processing parameters and benefits of enhanced sintering by uranium carbide liquid-phase sintering were established for the rapid and effective consolidation and formation of a solid-solution mixed carbide nuclear fuel.

Current status of U{sub 3}Si{sub 2} fuel element fabrication in Brazil

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

Durazzo, M.; Carvalho, E.F. Urano de; Saliba-Silva, A.M.

2008-07-15

IPEN has been working for increasing radioisotope production in order to supply the expanding demand for radiopharmaceutical medicines requested by the Brazilian welfare. To reach this objective, the IEA-R1 research reactor power capacity was recently increased from 2 MW to 4 MW. Since 1988 IPEN has been manufacturing its own fuel element, initially based on U{sub 3}O{sub 8}-Al dispersion fuel plates with 2.3 gU/cm{sup 3}. To support the reactor power increase, higher uranium density in the fuel plate meat had to be achieved for better irradiation flux and also to minimize the irradiated fuel elements to be stored. Uranium silicidemore » was the chosen option and the fuel fabrication development started with the support of the IAEA BRA/4/047 Technical Cooperation Project. This paper describes the results of this program and the current status of silicide fuel fabrication and its qualification. (author)« less

Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding

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

Robinson, Sharon M.; Chattin, Marc Rhea; Giaquinto, Joseph

2015-09-01

It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels). To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritium labelled water (THO) or T 2O. In a standard processing flowsheet, tritium management would bemore » accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy® cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste for Zircaloy® clad fuel by recovering the zirconium from the cladding for reuse. These recycle processes could release the tritium in the cladding. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 0–96%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. Samples of Surry-2 and H.B. Robinson pressurized water reactor cladding were heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. Cladding samples were also heated within the temperature range of 480–600ºC expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. The tritium content of the Surry-2 and H.B. Robinson cladding was measured to be ~234 and ~500 µCi/g, respectively. Heating the Surry-2 cladding at 500°C for 24 h removed ~0.2% of the tritium from the cladding, and heating at 700°C for 24 h removed ~9%. Heating the H.B. Robinson cladding at 700°C for 24 h removed ~11% of the tritium. When samples of the Surry-2 and H.B. Robinson claddings were heated at 700°C for 96 h, essentially all of the tritium in the cladding was removed. However, only ~3% of the tritium was removed when a sample of Surry-2 cladding was heated at 600°C for 96 h. These data indicate that the amount of tritium released from tritium pretreatment systems will be dependent on both the operating temperature and length of time in the system. Under certain conditions, a significant fraction of the tritium could remain bound in the cladding and would need to be considered in operations involving cladding recycle.« less

Application of low-frequency eddy current testing to the inspection of a double-walled tank in a reprocessing plant

NASA Astrophysics Data System (ADS)

Tajima, Naoki; Yusa, Noritaka; Hashizume, Hidetoshi

2018-04-01

This paper discusses the applicability of simple low-frequency eddy current testing to the detection of deeply embedded flaws. The study specifically considered a double tank in a reprocessing plant for extracting plutonium-uranium from spent nuclear fuels. The tank was modelled by two type 304 austenitic stainless steel plates situated with an air gap of 80 mm, and the change in the thickness of one of the plates was detected through the other plate and the air gap. Axisymmetric two-dimensional finite element simulations were conducted and found that a simple circular coil with a large diameter enabled to detect the thickness based on the magnetic flux density at the centre of the coil although the plates were as thick as 30 mm. The results of the numerical simulations were validated by experiments.

A Brief User's Guide to the Excel ® -Based DF Calculator

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

Jubin, Robert T.

2016-06-01

To understand the importance of capturing penetrating forms of iodine as well as the other volatile radionuclides, a calculation tool was developed in the form of an Excel ® spreadsheet to estimate the overall plant decontamination factor (DF). The tool requires the user to estimate splits of the volatile radionuclides within the major portions of the reprocessing plant, speciation of iodine and individual DFs for each off-gas stream within the Used Nuclear Fuel reprocessing plant. The Impact to the overall plant DF for each volatile radionuclide is then calculated by the tool based on the specific user choices. The Excelmore » ® spreadsheet tracks both elemental and penetrating forms of iodine separately and allows changes in the speciation of iodine at each processing step. It also tracks 3H, 14C and 85Kr. This document provides a basic user's guide to the manipulation of this tool.« less

Post-irradiation examination of prototype Al-64 wt% U{sub 3}Si{sub 2} fuel rods from NRU

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

Sears, D.F.; Primeau, M.F.; Buchanan, C.

1997-08-01

Three prototype fuel rods containing Al-64 wt% U{sub 3}Si{sub 2} (3.15 gU/cm{sup 3}) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U{sub 3}Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U{sub 3}Si{sub 2} powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37{degrees}C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinationsmore » showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U{sub 3}Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL`s research reactors.« less

Heterogeneous sodium fast reactor designed for transmuting minor actinide waste isotopes into plutonium fuel

NASA Astrophysics Data System (ADS)

Bays, Samuel Eugene

2008-10-01

In the past several years there has been a renewed interest in sodium fast reactor (SFR) technology for the purpose of destroying transuranic waste (TRU) produced by light water reactors (LWR). The utility of SFRs as waste burners is due to the fact that higher neutron energies allow all of the actinides, including the minor actinides (MA), to contribute to fission. It is well understood that many of the design issues of LWR spent nuclear fuel (SNF) disposal in a geologic repository are linked to MAs. Because the probability of fission for essentially all the "non-fissile" MAs is nearly zero at low neutron energies, these isotopes act as a neutron capture sink in most thermal reactor systems. Furthermore, because most of the isotopes produced by these capture reactions are also non-fissile, they too are neutron sinks in most thermal reactor systems. Conversely, with high neutron energies, the MAs can produce neutrons by fast fission. Additionally, capture reactions transmute the MAs into mostly plutonium isotopes, which can fission more readily at any energy. The transmutation of non-fissile into fissile atoms is the premise of the plutonium breeder reactor. In a breeder reactor, not only does the non-fissile "fertile" U-238 atom contribute fast fission neutrons, but also transmutes into fissile Pu-239. The fissile value of the plutonium produced by MA transmutation can only be realized in fast neutron spectra. This is due to the fact that the predominate isotope produced by MA transmutation, Pu-238, is itself not fissile. However, the Pu-238 fission cross section is significantly larger than the original transmutation parent, predominately: Np-237 and Am-241, in the fast energy range. Also, Pu-238's fission cross section and fission-to-capture ratio is almost as high as that of fissile Pu-239 in the fast neutron spectrum. It is also important to note that a neutron absorption in Pu-238, that does not cause fission, will instead produce fissile Pu-239. Given this fast fissile quality and also the fact that Pu-238 is transmuted from Np-237 and Am-241, these MAs are regarded as fertile material in the SFR design proposed by this dissertation. This dissertation demonstrates a SFR design which is dedicated to plutonium breeding by targeting Am-241 transmutation. This SFR design uses a moderated axial transmutation target that functions primarily as a pseudo-blanket fuel, which is reprocessed with the active driver fuel in an integrated recycling strategy. This work demonstrates the cost and feasibility advantages of plutonium breeding via MA transmutation by adopting reactor, reprocessing and fuel technologies previously demonstrated for traditional breeder reactors. The fuel cycle proposed seeks to find a harmony between the waste management advantages of transuranic burning SFRs and the resource sustainability of traditional plutonium breeder SFRs. As a result, the enhanced plutonium conversion from MAs decreases the burner SFR's fuel costs, by extracting more fissile value from the initial TRU purchased through SNF reprocessing.

Estimation of carbon 14 inventory in hull and end-piece wastes from Japanese commercial reprocessing operation

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

Tomofumi Sakuragi; Hiromi Tanabe; Emiko Hirose

2013-07-01

Hull and end-piece wastes generated from reprocessing plant operations are expected to be disposed of in a deep underground repository as Group 2 TRU wastes under the Japanese classification system. The activated metals that compose the spent fuel assemblies such as Zircaloy claddings and stainless steel nozzles are mixed and compressed after fuel dissolution, and then stuffed into stainless steel canisters. Carbon 14 is a typical activated product in the hulls and end-pieces and is mainly generated by the {sup 14}N(n,p){sup 14}C reaction. In the previous safety assessment of the TRU waste in Japan, the radionuclides inventory was calculated bymore » ORIGEN-2 code. Some conservative assumptions and preliminary estimates were used in this calculation. For example, total radionuclides generated from a single type of fuel assembly (45 GWd/tU for a PWR unit), and the thickness of the Zircaloy oxide film on the hulls (80 μm) were both overestimated. The second assumption in particular has a large effect on exposure dose evaluation. Therefore, it is essential to have a realistic source term evaluation regarding such items as the C-14 inventory and its distribution to waste parts. In the present study, a C-14 inventory of the hull and end-piece wastes from the operation of a commercial reprocessing plant in Japan corresponding to 32,000 tU (16,000 tU in each BWR and PWR) was calculated. Analysis using individual irradiation conditions and fuel characteristics was conducted on 6 types of fuel assemblies for BWRs and 12 types for PWRs (4 pile types x 3 burnup limits). The oxide film thickness data for each fuel type cladding were obtained from the published literature. Activation calculations were performed by using ORIGEN-2 code. For the amount of spent assembly and other waste characteristics, representative values were assumed based on the published literature. As a preliminary experiment, C-14 in irradiated BWR claddings was measured and found to be consistent with the calculated activation. The total C-14 inventory was estimated as 4.46x10{sup 14} Bq, consisting of 2.58x10{sup 14} Bq for BWRs and 1.87x10{sup 14} Bq for PWRs, and is consistent with the safety assessment of 4.4x10{sup 14} Bq. However, the distribution of the C-14 inventory to hull oxide, which was estimated under the assumption of instantaneous radionuclide release in the safety assessment, decreased from 5.72x10{sup 13} Bq (13% of the total) in the previous assessment to 1.30x10{sup 13} Bq (2.9% of the total; consisting of 1.48x10{sup 12} for BWRs and 1.15x10{sup 13} for PWRs). In other words, the exposure dose peak is reduced to approximate 25% of its previous value due to the use of detailed oxide film data that the BWR cladding has a thin oxide film. Other instantaneous release components for C-14 such as the fuel residual were negligible. (authors)« less

Recycling of paper: accounting of greenhouse gases and global warming contributions.

PubMed

Merrild, Hanna; Damgaard, Anders; Christensen, Thomas H

2009-11-01

Greenhouse gas (GHG) emissions have been established for recycling of paper waste with focus on a material recovery facility (MRF). The MRF upgrades the paper and cardboard waste before it is delivered to other industries where new paper or board products are produced. The accounting showed that the GHG contributions from the upstream activities and operational activities, with global warming factors (GWFs) of respectively 1 to 29 and 3 to 9 kg CO(2)-eq. tonne(- 1) paper waste, were small in comparison wih the downstream activities. The GHG contributions from the downstream reprocessing of the paper waste ranged from approximately 490 to 1460 kg CO(2)-eq. tonne( -1) of paper waste. The system may be expanded to include crediting of avoided virgin paper production which would result in GHG contributions from -1270 to 390 kg CO(2)-eq. tonne(- 1) paper waste. It may also be assumed that the wood not used for virgin paper production instead is used for production of energy that in turn is assumed to substitute for fossil fuel energy. This would result in GHG contributions from -1850 to -4400 kg CO(2)-eq. tonne(- 1) of paper waste. These system expansions reveal very large GHG savings, suggesting that the indirect upstream and operational GHG contributions are negligible in comparison with the indirect downstream emissions. However, the data for reprocessing of paper waste and the data for virgin paper production are highly variable. These differences are mainly related to different energy sources for the mills, both in regards to energy form (heat or electricity) and fuel (biomass or fossil fuels).

USHPRR FUEL FABRICATION PILLAR: FABRICATION STATUS, PROCESS OPTIMIZATIONS, AND FUTURE PLANS

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

Wight, Jared M.; Joshi, Vineet V.; Lavender, Curt A.

The Fuel Fabrication (FF) Pillar, a project within the U.S. High Performance Research Reactor Conversion program of the National Nuclear Security Administration’s Office of Material Management and Minimization, is tasked with the scale-up and commercialization of high-density monolithic U-Mo fuel for the conversion of appropriate research reactors to use of low-enriched fuel. The FF Pillar has made significant steps to demonstrate and optimize the baseline co-rolling process using commercial-scale equipment at both the Y-12 National Security Complex (Y-12) and BWX Technologies (BWXT). These demonstrations include the fabrication of the next irradiation experiment, Mini-Plate 1 (MP-1), and casting optimizations at Y-12.more » The FF Pillar uses a detailed process flow diagram to identify potential gaps in processing knowledge or demonstration, which helps direct the strategic research agenda of the FF Pillar. This paper describes the significant progress made toward understanding the fuel characteristics, and models developed to make informed decisions, increase process yield, and decrease lifecycle waste and costs.« less

Radioactivity near the sunken submarine "Kursk" in the Southern Barents Sea.

PubMed

Matishov, Genady G; Matishov, Dimitry G; Namiatov, Alexey E; Smith, John Norton; Carroll, Jolynn; Dahle, Salve

2002-05-01

Radioactivity measurements were conducted on seawater, sediment, and biota samples collected in the vicinity of the Russian submarine "Kursk" in September, 2000, within 1 month of the vessel's sinking in the Barents Sea to determine whether leakage of radioactivity from the vessel's two nuclear reactors had occurred and to assess the impact on one of the most productive fishing areas in the world. Levels of radioactivity in surface sediments and biota are within the range of values previously measured in the Barents Sea and can be ascribed to inputs from global fallout, European nuclear fuel reprocessing facilities, and the Chernobyl accident. However, levels of 1291 in seawater in the Southern Barents Sea increased by 500% between 1992 and 2000, and the 129I/137Cs ratio increased by more than an order of magnitude during this time, owing to long-range transport of releases from reprocessing facilities at Sellafield (U.K.) and La Hague (France). Although these results indicate that, at the time of sampling, leakage from the Kursk had a negligible impact on the environment, they also show that regional background levels of artificial radioactivity are varying rapidly on annual timescales and that Europe's nuclear reprocessing facilities are the leading contributor of anthropogenic radioactivity to the region.

[Micro fabricated enzyme battery].

PubMed

Sasaki, S; Karube, I

1996-10-01

Although various work has been done in the field of implantable micro actuators such as artificial organs and micro surgery robots, a suitable electric power supply for these is yet to be developed. For this purpose a micro fabricated enzyme fuel cell was developed which uses glucose contained in the human body as a fuel. In order to obtain enough voltage each cell was formed as part of a serial array on a silicon wafer. Glucose solution enters the cells by a capillary effect. In this article fuel cells already developed using biocatalysts are described, and the future possibility of a micro fabricated enzyme battery is discussed.

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

Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.

This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. Thesemore » glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability. (C) 2017 Published by Elsevier B.V.« less

Studies and research concerning BNFP: computerized nuclear materials control and accounting system development evaluation report, FY 1978

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

Crawford, J M; Ehinger, M H; Joseph, C

1978-10-01

Development work on a computerized system for nuclear materials control and accounting in a nuclear fuel reprocessing plant is described and evaluated. Hardware and software were installed and tested to demonstrate key measurement, measurement control, and accounting requirements at accountability input/output points using natural uranium. The demonstration included a remote data acquisition system which interfaces process and special instrumentation to a cenral processing unit.

Off-Site Monitoring of Nuclear Fuel Reprocessing Plants for Nuclear Weapons Proliferation

DTIC Science & Technology

1980-01-01

of commercial nuclear power reactors by the collection of cesium and neodynium radionuclides and the use-of isotopic correlation techniques.Both...Both Goodwin (ref 1) and Clark (ref 2) investigated off-site monitoring of commercial nuclear power reactoze by the collection of cesium and neodynium...manner than that which is used for power production.Economical generation of electrical power requires a long sus- tained fission cycle whereas Pu-239

METHOD OF DISSOLVING REFRACTORY ALLOYS

DOEpatents

Helton, D.M.; Savolainen, J.K.

1963-04-23

This patent relates to the dissolution of alloys of uranium with zirconium, thorium, molybdenum, or niobium. The alloy is contacted with an anhydrous solution of mercuric chloride in a low-molecular-weight monohydric alcohol to produce a mercury-containing alcohol slurry. The slurry is then converted to an aqueous system by adding water and driving off the alcohol. The resulting aqueous slurry is electrolyzed in the presence of a mercury cathode to remove the mercury and produce a uranium-bearing aqueous solution. This process is useful for dissolving irradiated nuclear reactor fuels for radiochemical reprocessing by solvent extraction. In addition, zirconium-alloy cladding is selectively removed from uranium dioxide fuel compacts by this means. (AEC)

Site remediation techniques in India: a review

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

Anomitra Banerjee; Miller Jothi

India is one of the developing countries operating site remediation techniques for the entire nuclear fuel cycle waste for the last three decades. In this paper we intend to provide an overview of remediation methods currently utilized at various hazardous waste sites in India, their advantages and disadvantages. Over the years the site remediation techniques have been well characterized and different processes for treatment, conditioning and disposal are being practiced. Remediation Methods categorized as biological, chemical or physical are summarized for contaminated soils and environmental waters. This paper covers the site remediation techniques implemented for treatment and conditioning of wastelandsmore » arising from the operation of nuclear power plant, research reactors and fuel reprocessing units. (authors)« less

RERTR-9 Summary Report

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

D. M. Perez

2011-05-01

The RERTR-9 experiment was designed to test the effect of modified fuel/clad interfaces in monolithic fuel plates and to demonstrate that the addition of Si to the matrix material in dispersion plates continued to be effective at high loading (~8.5 g U/cc). Several monolithic fuel plates were fabricated by Hot Isostatic Pressing (HIP) and Friction Bonding (FB) with thin layers of Si inserted and by HIP with a Zr diffusion barrier between the fuel and cladding. Si was applied to the interface by thermal spray of Al Si mixtures and by the insertion of thin Si-rich Al alloy foil betweenmore » the fuel/clad interface. The dispersion fuel plates were fabricated by semi-standard rolling techniques (the reduction by rolling was lowered to limit fabrication defects). Matrix materials consisted of Al-Si alloys and mixtures with various levels of Si. The following report summarizes the life of the RERTR-9A/B experiment through end of irradiation, including as-run neutronic analysis, thermal analysis and hydraulic testing results.« less

Sources and transport of anthropogenic radionuclides in the Ob River system, Siberia

NASA Astrophysics Data System (ADS)

Cochran, J. Kirk; Moran, S. Bradley; Fisher, Nicholas S.; Beasley, Thomas M.; Kelley, James M.

2000-06-01

The potential sources of anthropogenic radionuclides to the Ob River system of western Siberia include global stratospheric fallout, tropospheric fallout from atomic weapons tests and releases from production and reprocessing facilities. Samples of water, suspended and bottom sediments collected in 1994 and 1995 have been used to characterize the sources and transport of 137Cs, Pu isotopes, 237Np and 129I through the system. For the radionuclides that associate with particles, isotope ratios provide clues to their sources, providing any geochemical fractionation can be taken into account. Activity ratios of 239,240Pu/ 137Cs in suspended sediments are lower than the global fallout ratio in the Irtysh River before its confluence with the Ob, comparable to fallout in the central reach of the Ob, and greater than the fallout values in the lower Ob and in the Taz River. This pattern mirrors the downriver decrease in dissolved organic carbon (DOC) concentrations. Laboratory adsorption experiments with Ob River sediment and water show that Kd values for Am (and presumably other actinides) are depressed by two orders of magnitude in the presence of Ob DOC concentrations, relative to values measured in DOC-free Ob water. Iodine and cesium Kd values show little or no (less than a factor of 2) dependence on DOC. Mixing plots using plutonium isotope ratios (atom ratios) show that Pu in suspended sediments of the Ob is a mixture of stratospheric global fallout at northern latitudes, tropospheric fallout from the former Soviet Union test site at Semipalatinsk and reprocessing of spent fuel at Tomsk-7. Plutonium from Semipalatinsk is evident in the Irtysh River above its confluence with the Tobal. Suspended sediment samples taken in the Ob above its confluence with the Irtysh indicate the presence of Pu derived from the Tomsk-7 reprocessing facilities. A mixing plot constructed using 237Np/ 239Pu vs. 240Pu/ 239Pu shows similar mixtures of stratospheric and tropospheric fallout, with the likely addition of inputs from reprocessing facilities and reactor operations. As with Pu/Cs ratios, Np/Pu ratios could be modified by differential geochemical behaviors of Np and Pu. Dissolved 129I only weakly interacts with particles in the Ob; size-fractionated sampling shows that the colloidal 129I fraction (defined as 1 kDa-0.2 μm) contains generally <5% relative to that passing a 0.2 μm filter. Iodine-129 concentrations decrease from 8.3×10 9 to 0.65×10 9 atoms l -1 through the Ob system toward the Kara Sea, with highest values in the Tobal River and lowest in the Taz River. The likely source of the elevated 129I in the Tobal is release from the production-reprocessing facilities at Mayak, and decreases downriver are predominantly due to dilution as the various tributaries with low 129I join the system. Fluxes of 129I to the lower Ob at Salekhard are <1% of the releases of this radionuclide from the nuclear fuel reprocessing facilities at Sellafield, UK and La Hague, France.

Fabrication of High Temperature Cermet Materials for Nuclear Thermal Propulsion

NASA Technical Reports Server (NTRS)

Hickman, Robert; Panda, Binayak; Shah, Sandeep

2005-01-01

Processing techniques are being developed to fabricate refractory metal and ceramic cermet materials for Nuclear Thermal Propulsion (NTP). Significant advances have been made in the area of high-temperature cermet fuel processing since RoverNERVA. Cermet materials offer several advantages such as retention of fission products and fuels, thermal shock resistance, hydrogen compatibility, high conductivity, and high strength. Recent NASA h d e d research has demonstrated the net shape fabrication of W-Re-HfC and other refractory metal and ceramic components that are similar to UN/W-Re cermet fuels. This effort is focused on basic research and characterization to identify the most promising compositions and processing techniques. A particular emphasis is being placed on low cost processes to fabricate near net shape parts of practical size. Several processing methods including Vacuum Plasma Spray (VPS) and conventional PM processes are being evaluated to fabricate material property samples and components. Surrogate W-Re/ZrN cermet fuel materials are being used to develop processing techniques for both coated and uncoated ceramic particles. After process optimization, depleted uranium-based cermets will be fabricated and tested to evaluate mechanical, thermal, and hot H2 erosion properties. This paper provides details on the current results of the project.

Development of High-Temperature Transport Technologies of Molten Salt Slurry in Pyrometallurgical Reprocessing

NASA Astrophysics Data System (ADS)

Hijikata, Takatoshi; Koyama, Tadafumi

Pyrometallurgical-reprocessing is one of the most promising technologies for advanced fuel cycle with favorable economic potential and intrinsic proliferation resistance. The development of transport technology for molten salt is a key issue in the industrialization of pyro-reprocessing. As for pure molten LiCl-KCl eutectic salt at approximately 773 K, we have already reported the successful results of transport using gravity and a centrifugal pump. However, molten salt in an electrorefiner mixes with insoluble fines when spent fuel is dissolved in porous anode basket. The insoluble consists of noble metal fission products, such as Pd, Ru, Mo, and Zr. There have been very few transport studies of a molten salt slurry (metal fines-molten salt mixture). Hence, transport experiments on a molten salt slurry were carried out to investigate the behavior of the slurry in a tube. The apparatus used in the transport experiments on the molten salt slurry consisted of a supply tank, a 10° inclined transport tube (10 mm inner diameter), a valve, a filter, and a recovery tank. Stainless steel (SS) fines with diameters from 53 to 415 μm were used. To disperse these fines homogenously, the molten salt and fines were stirred in the supply tank by an impeller at speeds from 1200 to 2100 rpm. The molten salt slurry containing 0.04 to 0.4 vol.% SS fines was transported from the supply tank to the recovery tank through the transportation tube. In the recovery tank, the fines were separated from the molten salt by the filter to measure the transport behavior of molten salt and SS fines. When the velocity of the slurry was 0.02 m/s, only 1% of the fines were transported to the recovery tank. On the other hand, most of the fines were transported when the velocity of the slurry was more than 0.8 m/s. Consequently, the molten salt slurry can be transported when the velocity is more than 0.8 m/s.

An interface tracking model for droplet electrocoalescence.

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

Erickson, Lindsay Crowl

This report describes an Early Career Laboratory Directed Research and Development (LDRD) project to develop an interface tracking model for droplet electrocoalescence. Many fluid-based technologies rely on electrical fields to control the motion of droplets, e.g. microfluidic devices for high-speed droplet sorting, solution separation for chemical detectors, and purification of biodiesel fuel. Precise control over droplets is crucial to these applications. However, electric fields can induce complex and unpredictable fluid dynamics. Recent experiments (Ristenpart et al. 2009) have demonstrated that oppositely charged droplets bounce rather than coalesce in the presence of strong electric fields. A transient aqueous bridge forms betweenmore » approaching drops prior to pinch-off. This observation applies to many types of fluids, but neither theory nor experiments have been able to offer a satisfactory explanation. Analytic hydrodynamic approximations for interfaces become invalid near coalescence, and therefore detailed numerical simulations are necessary. This is a computationally challenging problem that involves tracking a moving interface and solving complex multi-physics and multi-scale dynamics, which are beyond the capabilities of most state-of-the-art simulations. An interface-tracking model for electro-coalescence can provide a new perspective to a variety of applications in which interfacial physics are coupled with electrodynamics, including electro-osmosis, fabrication of microelectronics, fuel atomization, oil dehydration, nuclear waste reprocessing and solution separation for chemical detectors. We present a conformal decomposition finite element (CDFEM) interface-tracking method for the electrohydrodynamics of two-phase flow to demonstrate electro-coalescence. CDFEM is a sharp interface method that decomposes elements along fluid-fluid boundaries and uses a level set function to represent the interface.« less

Design Evolutuion of Hot Isotatic Press Cans for NTP Cermet Fuel Fabrication

NASA Technical Reports Server (NTRS)

Mireles, O. R.; Broadway, J.; Hickman, R.

2014-01-01

Nuclear Thermal Propulsion (NTP) is under consideration for potential use in deep space exploration missions due to desirable performance properties such as a high specific impulse (> 850 seconds). Tungsten (W)-60vol%UO2 cermet fuel elements are under development, with efforts emphasizing fabrication, performance testing and process optimization to meet NTP service life requirements [1]. Fuel elements incorporate design features that provide redundant protection from crack initiation, crack propagation potentially resulting in hot hydrogen (H2) reduction of UO2 kernels. Fuel erosion and fission product retention barriers include W coated UO2 fuel kernels, W clad internal flow channels and fuel element external W clad resulting in a fully encapsulated fuel element design as shown.

Fabrication of low-temperature solid oxide fuel cells with a nanothin protective layer by atomic layer deposition

PubMed Central

2013-01-01

Anode aluminum oxide-supported thin-film fuel cells having a sub-500-nm-thick bilayered electrolyte comprising a gadolinium-doped ceria (GDC) layer and an yttria-stabilized zirconia (YSZ) layer were fabricated and electrochemically characterized in order to investigate the effect of the YSZ protective layer. The highly dense and thin YSZ layer acted as a blockage against electron and oxygen permeation between the anode and GDC electrolyte. Dense GDC and YSZ thin films were fabricated using radio frequency sputtering and atomic layer deposition techniques, respectively. The resulting bilayered thin-film fuel cell generated a significantly higher open circuit voltage of approximately 1.07 V compared with a thin-film fuel cell with a single-layered GDC electrolyte (approximately 0.3 V). PMID:23342963

Development of an alkaline fuel cell subsystem

NASA Technical Reports Server (NTRS)

1987-01-01

A two task program was initiated to develop advanced fuel cell components which could be assembled into an alkaline power section for the Space Station Prototype (SSP) fuel cell subsystem. The first task was to establish a preliminary SSP power section design to be representative of the 200 cell Space Station power section. The second task was to conduct tooling and fabrication trials and fabrication of selected cell stack components. A lightweight, reliable cell stack design suitable for the SSP regenerative fuel cell power plant was completed. The design meets NASA's preliminary requirements for future multikilowatt Space Station missions. Cell stack component fabrication and tooling trials demonstrated cell components of the SSP stack design of the 1.0 sq ft area can be manufactured using techniques and methods previously evaluated and developed.

Methods for making a porous nuclear fuel element

DOEpatents

Youchison, Dennis L; Williams, Brian E; Benander, Robert E

2014-12-30

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Human capital needs - teaching, training and coordination for nuclear fuel cycle

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

Retegan, T.; Ekberg, C.; John, J.

Human capital is the accumulation of competencies, knowledge, social and creativity skills and personality attributes, which are necessary to perform work so as to produce economic value. In the frame of the nuclear fuel cycle, this is of paramount importance that the right human capital exists and in Europe this is fostered by a series of integrated or directed projects. The teaching, training and coordination will be discussed in the frame of University curricula with examples from several programs, like e.g. the Master of Nuclear Engineering at Chalmers University, Sweden and two FP7 EURATOM Projects: CINCH - a project formore » cooperation in nuclear chemistry - and ASGARD - a research project on advanced or novel nuclear fuels and their reprocessing issues for generation IV reactors. The integration of the university curricula in the market needs but also the anchoring in the research and future fuel cycles will be also discussed, with examples from the ASGARD project. (authors)« less

FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Loeb, E.; Nicklas, J.H.

1959-02-01

A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

2012-01-01

Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes these overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

NASA Technical Reports Server (NTRS)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

2012-01-01

Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes thses overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

Dissolver vessel bottom assembly

DOEpatents

Kilian, Douglas C.

1976-01-01

An improved bottom assembly is provided for a nuclear reactor fuel reprocessing dissolver vessel wherein fuel elements are dissolved as the initial step in recovering fissile material from spent fuel rods. A shock-absorbing crash plate with a convex upper surface is disposed at the bottom of the dissolver vessel so as to provide an annular space between the crash plate and the dissolver vessel wall. A sparging ring is disposed within the annular space to enable a fluid discharged from the sparging ring to agitate the solids which deposit on the bottom of the dissolver vessel and accumulate in the annular space. An inlet tangential to the annular space permits a fluid pumped into the annular space through the inlet to flush these solids from the dissolver vessel through tangential outlets oppositely facing the inlet. The sparging ring is protected against damage from the impact of fuel elements being charged to the dissolver vessel by making the crash plate of such a diameter that the width of the annular space between the crash plate and the vessel wall is less than the diameter of the fuel elements.

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

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

Renfro, David G; Chandler, David; Cook, David Howard

2014-11-01

Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully convertedmore » using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.« less

FLOOR PLAN OF MAIN PROCESSING BUILDING (CPP601), SECOND FLOOR SHOWING ...

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

FLOOR PLAN OF MAIN PROCESSING BUILDING (CPP-601), SECOND FLOOR SHOWING PROCESS MAKEUP AREA AND EIGHTEEN CELLS AND ADJOINING REMOTE ANALYTICAL FACILITY (CPP-627) SHOWING COLD LAB, DECONTAMINATION ROOM, MULTICURIE CELL ROOM, AND OFFICES. TO LEFT ARE LABORATORY BUILDING (CPP-602) AND MAINTENANCE BUILDING (CPP-630). INL DRAWING NUMBER 200-0601-00-706-051980. ALTERNATE ID NUMBER CPP-E-1980. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

FLOOR PLAN OF MAIN PROCESSING BUILDING (CPP601), FIRST FLOOR SHOWING ...

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

FLOOR PLAN OF MAIN PROCESSING BUILDING (CPP-601), FIRST FLOOR SHOWING SAMPLE CORRIDORS AND EIGHTEEN CELLS AND ADJOINING REMOTE ANALYTICAL FACILITY (CPP-627) SHOWING REMOTE ANALYTICAL FACILITIES LAB, DECONTAMINATION ROOM, AND MULTICURIE CELL ROOM. TO LEFT ARE LABORATORY BUILDING (CPP-602) AND MAINTENANCE BUILDING (CPP-630). INL DRAWING NUMBER 200-0601-00-706-051979. ALTERNATE ID NUMBER CPP-E-1979. - Idaho National Engineering Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex, Scoville, Butte County, ID

Fundamental and applied aspects of the chemistry of radioactive iodine in gas and aqueous media

NASA Astrophysics Data System (ADS)

Kulyukhin, Sergei A.

2012-10-01

The results of investigations carried out in the past 15-20 years in the chemistry of radioactive iodine in the gas phase and in aqueous media of various compositions are described systematically and analyzed. The prospects for practical application of the obtained data in various fields of nuclear power industry including accidents at nuclear power plants and enterprises engaged in spent nuclear fuel reprocessing are estimated. The bibliography includes 206 references.

Monitoring Iodine-129 in Air and Milk Samples Collected Near the Hanford Site: An Investigation of Historical Iodine Monitoring Data

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

Fritz, Brad G.; Patton, Gregory W.

2006-01-01

While other research has reported on the concentrations of 129I in the environment surrounding active nuclear fuel reprocessing facilities, there is a shortage of information regarding how the concentrations change once facilities close. At the Hanford Site, the Plutonium-Uranium Extraction (PUREX) chemical separation plant was operational between 1983 and 1990, during which time 129I concentrations in air and milk were measured. After the cessation of operations in 1990, plant emissions decreased 2.5 orders of magnitude over an 8 year period, and monitoring of environmental levels continued. An evaluation of air and milk 129I concentration data spanning the PUREX operation andmore » post closure period was conducted to compare the changes in environmental levels of 129I measured. Measured concentrations over the monitoring period were below levels that could result in a potential human dose greater than 10 uSv. There was a significant and measurable difference in the measured air concentrations of 129I at different distances from the source, indicating a distinct Hanford fingerprint. Correlations between stack emissions of 129I and concentrations in air and milk indicate that atmospheric emissions were responsible for the 129I concentrations measured in environmental samples. The measured concentrations during PUREX operation were similar to observations made around a fuel reprocessing plant in Germany.« less

Spectroscopic methods of process monitoring for safeguards of used nuclear fuel separations

NASA Astrophysics Data System (ADS)

Warburton, Jamie Lee

To support the demonstration of a more proliferation-resistant nuclear fuel processing plant, techniques and instrumentation to allow the real-time, online determination of special nuclear material concentrations in-process must be developed. An ideal materials accountability technique for proliferation resistance should provide nondestructive, realtime, on-line information of metal and ligand concentrations in separations streams without perturbing the process. UV-Visible spectroscopy can be adapted for this precise purpose in solvent extraction-based separations. The primary goal of this project is to understand fundamental URanium EXtraction (UREX) and Plutonium-URanium EXtraction (PUREX) reprocessing chemistry and corresponding UV-Visible spectroscopy for application in process monitoring for safeguards. By evaluating the impact of process conditions, such as acid concentration, metal concentration and flow rate, on the sensitivity of the UV-Visible detection system, the process-monitoring concept is developed from an advanced application of fundamental spectroscopy. Systematic benchtop-scale studies investigated the system relevant to UREX or PUREX type reprocessing systems, encompassing 0.01-1.26 M U and 0.01-8 M HNO3. A laboratory-scale TRansUranic Extraction (TRUEX) demonstration was performed and used both to analyze for potential online monitoring opportunities in the TRUEX process, and to provide the foundation for building and demonstrating a laboratory-scale UREX demonstration. The secondary goal of the project is to simulate a diversion scenario in UREX and successfully detect changes in metal concentration and solution chemistry in a counter current contactor system with a UV-Visible spectroscopic process monitor. UREX uses the same basic solvent extraction flowsheet as PUREX, but has a lower acid concentration throughout and adds acetohydroxamic acid (AHA) as a complexant/reductant to the feed solution to prevent the extraction of Pu. By examining UV-Visible spectra gathered in real time, the objective is to detect the conversion from the UREX process, which does not separate Pu, to the PUREX process, which yields a purified Pu product. The change in process chemistry can be detected in the feed solution, aqueous product or in the raffinate stream by identifying the acid concentration, metal distribution and the presence or absence of AHA. A fiber optic dip probe for UV-Visible spectroscopy was integrated into a bank of three counter-current centrifugal contactors to demonstrate the online process monitoring concept. Nd, Fe and Zr were added to the uranyl nitrate system to explore spectroscopic interferences and identify additional species as candidates for online monitoring. This milestone is a demonstration of the potential of this technique, which lies in the ability to simultaneously and directly monitor the chemical process conditions in a reprocessing plant, providing inspectors with another tool to detect nuclear material diversion attempts. Lastly, dry processing of used nuclear fuel is often used as a head-end step before solvent extraction-based separations such as UREX or TRUEX. A non-aqueous process, used fuel treatment by dry processing generally includes chopping of used fuel rods followed by repeated oxidation-reduction cycles and physical separation of the used fuel from the cladding. Thus, dry processing techniques are investigated and opportunities for online monitoring are proposed for continuation of this work in future studies.

Fukushima Daiichi Information Repository FY13 Status

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

Smith, Curtis; Phelan, Cherie; Schwieder, Dave

The accident at the Fukushima Daiichi nuclear power station in Japan is one of the most serious in commercial nuclear power plant operating history. Much will be learned that may be applicable to the U.S. reactor fleet, nuclear fuel cycle facilities, and supporting systems, and the international reactor fleet. For example, lessons from Fukushima Daiichi may be applied to emergency response planning, reactor operator training, accident scenario modeling, human factors engineering, radiation protection, and accident mitigation; as well as influence U.S. policies towards the nuclear fuel cycle including power generation, and spent fuel storage, reprocessing, and disposal. This document describesmore » the database used to establish a centralized information repository to store and manage the Fukushima data that has been gathered. The data is stored in a secured (password protected and encrypted) repository that is searchable and available to researchers at diverse locations.« less

Summary of the Forty-Fifth NCRP annual meeting on "the future of nuclear power worldwide: safety, health and the environment".

PubMed

Corradini, Michael

2011-01-01

The role of nuclear power as a major resource in meeting the projected growth of electric power requirements in the United States and worldwide during the 21st century is a subject of great contemporary interest. The goal of the 2009 NCRP Annual Meeting was to provide a forum for an in-depth discussion of issues related to the safety, health and environmental protection aspects of new nuclear power reactor systems and related fuel-cycle facilities such as fuel production and reprocessing strategies. The meeting was an international conference with participation of almost 400 representatives from many nations, scientific organizations, nuclear industries, and governmental agencies engaged in the development and regulatory control of advanced nuclear reactor systems and fuel-cycle operations. Highlights of the meeting are summarized in this report. Copyright © 2010 Health Physics Society

Dangers associated with civil nuclear power programmes: weaponization and nuclear waste.

PubMed

Boulton, Frank

2015-07-24

The number of nuclear power plants in the world rose exponentially to 420 by 1990 and peaked at 438 in 2002; but by 2014, as closed plants were not replaced, there were just 388. In spite of using more renewable energy, the world still relies on fossil fuels, but some countries plan to develop new nuclear programmes. Spent nuclear fuel, one of the most dangerous and toxic materials known, can be reprocessed into fresh fuel or into weapons-grade materials, and generates large amounts of highly active waste. This article reviews available literature on government and industry websites and from independent analysts on world energy production, the aspirations of the 'new nuclear build' programmes in China and the UK, and the difficulties in keeping the environment safe over an immense timescale while minimizing adverse health impacts and production of greenhouse gases, and preventing weaponization by non-nuclear-weapons states acquiring civil nuclear technology.

Thermally and chemically responsive nanoporous materials for efficient capture of fission product gases.

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

Stroeve, Pieter; Faller, Roland

The objective of this project was to develop robust, high-efficiency materials for capture of fission product gases such as He, Xe and Kr in scenarios relevant for both reactor fuels and reprocessing operations. The relevant environments are extremely harsh, encompassing temperatures up to 1500 °C, high levels of radiation, as well as potential exposures to highly-reactive chemicals such as nitric acid and organic solvents such as kerosene. The requirement for nanostructured capture materials is driven in part by the very short (few micron) diffusion distances for product gases in nuclear fuel. We achieved synthesis, characterization and detailed modeling of themore » materials. Although not all materials reviewed in this report will be feasible for the ultimate goal of integration in nuclear fuel, nevertheless each material studied has particular properties which will enable an optimized material to be efficiently developed and characterized.« less

The nuclear energy outlook--a new book from the OECD nuclear energy agency.

PubMed

Yoshimura, Uichiro

2011-01-01

This paper summarizes the key points of a report titled Nuclear Energy Outlook, published in 2008 by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, which has 30 member nations. The report discusses the commitment of many nations to increase nuclear power generating capacity and the potential rate of building new electricity-generating nuclear plants by 2030 to 2050. The resulting decrease in carbon dioxide emissions from fossil fuel combustion resulting from an increase in nuclear power sources is described. Other topics that are discussed include the need to develop non-proliferative nuclear fuels, the importance of developing geological disposal facilities or reprocessing capabilities for spent nuclear fuel and high-level radioactive waste materials, and the requirements for a larger nuclear workforce and greater cost competitiveness for nuclear power generation. Copyright © 2010 Health Physics Society

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.

Recent GE BWR fuel experience and design evolution

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

Wood, J.E.; Potts, G.A.; Proebstle, R.A.

1992-01-01

Reliable fuel operation is essential to the safe, reliable, and economic power production by today's commercial nuclear reactors. GE Nuclear Energy is committed to maximize fuel reliability through the progressive development of improved fuel design features and dedication to provide the maximum quality of the design features and dedication to provide the maximum quality of the design, fabrication, and operation of GE BWR fuel. Over the last 35 years, GE has designed, fabricated, and placed in operation over 82,000 BWR fuel bundles containing over 5 million fuel rods. This experience includes successful commercial reactor operation of fuel assemblies to greatermore » than 45000 MWd/MTU bundle average exposure. This paper reports that this extensive experience base has enabled clear identification and characterization of the active failure mechanisms. With this failure mechanism characterization, mitigating actions have been developed and implemented by GE to provide the highest reliability BWR fuel bundles possible.« less

Fully Ceramic Microencapsulated Fuel Development for LWR Applications

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

Snead, Lance Lewis; Besmann, Theodore M; Terrani, Kurt A

2012-01-01

The concept, fabrication, and key feasibility issues of a new fuel form based on the microencapsulated (TRISO-type) fuel which has been specifically engineered for LWR application and compacted within a SiC matrix will be presented. This fuel, the so-called fully ceramic microencapsulated fuel is currently undergoing development as an accident tolerant fuel for potential UO2 replacement in commercial LWRs. While the ability of this fuel to facilitate normal LWR cycle performance is an ongoing effort within the program, this will not be a focus of this paper. Rather, key feasibility and performance aspects of the fuel will be presented includingmore » the ability to fabricate a LWR-specific TRISO, the need for and route to a high thermal conductivity and fully dense matrix that contains neutron poisons, and the performance of that matrix under irradiation and the interaction of the fuel with commercial zircaloy clad.« less

Paper-based membraneless hydrogen peroxide fuel cell prepared by micro-fabrication

NASA Astrophysics Data System (ADS)

Mousavi Ehteshami, Seyyed Mohsen; Asadnia, Mohsen; Tan, Swee Ngin; Chan, Siew Hwa

2016-01-01

A paper-based membraneless single-compartment hydrogen peroxide power source prepared by micro-electromechanical systems (MEMS) technology is reported. The cell utilizes hydrogen peroxide as both fuel and oxidant in a low volume cell fabricated on paper. The fabrication method used is a simple method where precise, small-sized patterns are produced which include the hydrophilic paper bounded by hydrophobic resin. Open circuit potentials of 0.61 V and 0.32 V are achieved for the cells fabricated with Prussian Blue as the cathode and aluminium/nickel as the anode materials, respectively. The power produced by the cells is 0.81 mW cm-2 at 0.26 V and 0.38 mW cm-2 at 0.14 V, respectively, even after the cell is bent or distorted. Such a fuel cell provides an easily fabricated, environmentally friendly, flexible and cost saving power source. The cell may be integrated within a self-sustained diagnostic system to provide the on-demand power for future bio-sensing applications.

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

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

Myers, C.W.; Giraud, K.M.

Newcomer countries expected to develop new nuclear power programs by 2030 are being encouraged by the International Atomic Energy Agency to explore the use of shared facilities for spent fuel storage and geologic disposal. Multinational underground nuclear parks (M-UNPs) are an option for sharing such facilities. Newcomer countries with suitable bedrock conditions could volunteer to host M-UNPs. M-UNPs would include back-end fuel cycle facilities, in open or closed fuel cycle configurations, with sufficient capacity to enable M-UNP host countries to provide for-fee waste management services to partner countries, and to manage waste from the M-UNP power reactors. M-UNP potential advantagesmore » include: the option for decades of spent fuel storage; fuel-cycle policy flexibility; increased proliferation resistance; high margin of physical security against attack; and high margin of containment capability in the event of beyond-design-basis accidents, thereby reducing the risk of Fukushima-like radiological contamination of surface lands. A hypothetical M-UNP in crystalline rock with facilities for small modular reactors, spent fuel storage, reprocessing, and geologic disposal is described using a room-and-pillar reference-design cavern. Underground construction cost is judged tractable through use of modern excavation technology and careful site selection. (authors)« 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

Simple fabrication of 12 μm thin nanocomposite fuel cell membranes by direct electrospinning and printing

NASA Astrophysics Data System (ADS)

Breitwieser, Matthias; Klose, Carolin; Klingele, Matthias; Hartmann, Armin; Erben, Johannes; Cho, Hyeongrae; Kerres, Jochen; Zengerle, Roland; Thiele, Simon

2017-01-01

Direct membrane deposition (DMD) was recently introduced as a novel polymer electrolyte membrane fabrication method. Here, this approach is extended to fabricate 12 μm thin nanocomposite fuel cell membranes. Poly(vinylidene fluoride-co-hexafluoropropylene) (PVDF-HFP) nanofibers are directly electrospun onto gas diffusion electrodes. By inkjet-printing Nafion ionomer dispersion into the pore space of PVDF-HFP nanofiber mats, composite membranes of 12 μm thickness were fabricated. At 120 °C and 35% relative humidity, stoichiometric 1.5/2.5 H2/air flow and atmospheric pressure, the power density of the DMD fuel cell (0.19 W cm-2), was about 1.7 times higher than that of the reference fuel cell (0.11 W cm-2) with Nafion HP membrane and identical catalyst. A lower ionic resistance and, especially at 120 °C, a reduced charge transfer resistance is found compared to the Nafion HP membrane. A 100 h accelerated stress test revealed a voltage decay of below 0.8 mV h-1, which is in the range of literature values for significantly thicker reinforced membranes. Finally, this novel fabrication approach enables new degrees of freedom in the design of complex composite membranes. The presented combination of scalable deposition techniques has the potential to simplify and thus reduce cost of composite membrane fabrication at a larger scale.

Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

2011-03-01

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

DOEpatents

Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

2013-09-03

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

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

Bruffey, Stephanie H.; Jubin, Robert Thomas; Jordan, J. A.

U.S. regulations will require the removal of 129I from the off-gas streams of any used nuclear fuel (UNF) reprocessing plant prior to discharge of the off-gas to the environment. Multiple off-gas streams within a UNF reprocessing plant combine prior to release, and each of these streams contains some amount of iodine. For an aqueous UNF reprocessing plant, these streams include the dissolver off-gas, the cell off-gas, the vessel off-gas (VOG), the waste off-gas and the shear off-gas. To achieve regulatory compliance, treatment of multiple off-gas streams within the plant must be performed. Preliminary studies have been completed on the adsorptionmore » of I 2 onto silver mordenite (AgZ) from prototypical VOG streams. The study reported that AgZ did adsorb I 2 from a prototypical VOG stream, but process upsets resulted in an uneven feed stream concentration. The experiments described in this document both improve the characterization of I 2 adsorption by AgZ from dilute gas streams and further extend it to include characterization of the adsorption of organic iodides (in the form of CH 3I) onto AgZ under prototypical VOG conditions. The design of this extended duration testing was such that information about the rate of adsorption, the penetration of the iodine species, and the effect of sorbent aging on iodine removal in VOG conditions could be inferred.« less

Development of a conditioning system for the dual-purpose transport and storage cask for spent nuclear fuel from decommissioned Russian submarines

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

Dyer, R.S.; Barnes, E.; Snipes, R.L.

2007-07-01

Russia, stores large quantities of spent nuclear fuel (SNF) from submarine and ice-breaker nuclear powered naval vessels. This high-level radioactive material presents a significant threat to the Arctic and marine environments. Much of the SNF from decommissioned Russian nuclear submarines is stored either onboard the submarines or in floating storage vessels in Northwest and Far East Russia. Some of the SNF is damaged, stored in an unstable condition, or of a type that cannot currently be reprocessed. In many cases, the existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing all of this fuelmore » from remote locations. Additional transport and storage options are required. Some of the existing storage facilities being used in Russia do not meet health and safety and physical security requirements. The U.S. has assisted Russia in the development of a new dual-purpose metal-concrete transport and storage cask (TUK-108/1) for their military SNF and assisted them in building several new facilities for off-loading submarine SNF and storing these TUK-108/1 casks. These efforts have reduced the technical, ecological, and security challenges for removal, handling, interim storage, and shipment of this submarine fuel. Currently, Russian licensing limits the storage period of the TUK-108/1 casks to no more than two years before the fuel must be shipped for reprocessing. In order to extend this licensed storage period, a system is required to condition the casks by removing residual water and creating an inert storage environment by backfilling the internal canisters with a noble gas such as argon. The U.S. has assisted Russia in the development of a mobile cask conditioning system for the TUK-108/1 cask. This new conditioning system allows the TUK 108/1 casks to be stored for up to five years after which the license may be considered for renewal for an additional five years or the fuel will be shipped to 'Mayak' for reprocessing. The U.S. Environmental Protection Agency (EPA), in cooperation with the U.S. DOD Office of Cooperative Threat Reduction (CTR), and the DOE's ORNL, along with the Norwegian Defense Research Establishment, worked closely with the Ministry of Defense and the Ministry of Atomic Energy of the Russian Federation (RF) to develop an improved integrated management system for interim storage of military SNF in Russia. The initial Project activities included: (1) development of a prototype dual-purpose, metal-concrete 40-ton cask for both the transport and interim storage of RF SNF, and (2) development of the first transshipment/interim storage facility for these casks in Murmansk. The U.S. has continued support to the project by assisting the RF with the development of the first mobile system that provides internal conditioning for the TUK-108/1 casks to allow them to be stored for longer than the current licensing period of two years. Development of the prototype TUK-108/1 cask was completed in December 2000 under the Arctic Military Environmental Cooperation (AMEC) Program. This was the first metal-concrete cask developed, licensed, and produced in the RF for both the transportation and storage of SNF from decommissioned submarines. These casks are currently being serially produced in NW Russia and 108 casks have been produced to date. Russia is using these casks for the transport and interim storage of military SNF from decommissioned nuclear submarines at naval installations in the Arctic and Far East in conformance with the Strategic Arms Reduction Treaty (START II). The design, construction, and commissioning of the first transshipment/interim storage facility in the RF was completed and ready for full operation in September 2003. Because of the RF government reorganization and changing regulations for spent fuel storage facilities, the storage facility at Murmansk was not fully licensed for operation until December 2005. The RF has reported that the facility is now fully operational. The TUK-108/1 SNF transport and storage casks were designed to have a 50-year storage life. Current RF practice is not to condition the submarine SNF or cask during the cask loading. Current RF regulations allow up to 4 mm of residual water (up to 3.2 liters) to remain in the casks. It has been determined that allowing this amount of residual water to remain untreated for a period longer than two years can produce hydrogen gas through hydrolysis which will increase the risk of explosion and could cause some corrosion of internal components. A solution to this problem was to develop and utilize a cask conditioning system to remove the residual water and create an inert storage environment in the cask by back-filling the internal cask cavity with an inert gas, such as helium or argon. This system is compatible with the existing TUK-108/1 design and is mobile for use at multiple submarine dismantlement sites. The RF has required that this cask conditioning system be tested and commissioned at the 'Zvezda' Shipyard in the Far East near Vladivostok, one of the major RF submarine fuel off loading and storage facilities. Currently, the fuel cannot be transferred to 'Mayak' for reprocessing until the completion of the 20 km railroad connector between 'Zvezda' and the main rail line to 'Mayak'. The cask conditioning system will allow extension of the currently-stored casks for an additional three years, at which time the rail connector line should be completed. The current license to store these casks at 'Zvezda' was scheduled to expire on 31 Dec 2006. Without the cask-conditioning system, the license could not be extended, no more fuel could be off-loaded from the decommissioned submarines, and the START objectives could not be met at 'Zvezda'. Completion of this cask conditioning system has removed a significant bottleneck for the completion of the Russian submarine decommissioning program under the START II Agreement. (authors)« less

Iodine Adsorption in Metal Organic Frameworks in the Presence of Humidity

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

Banerjee, Debasis; Chen, Xianyin; Lobanov, Sergey S.

Used nuclear fuel (UNF) reprocessing represents a unique challenge when dealing with radionuclides such as isotopes of 85Kr and 129I 2, due to their volatility and long half-life. However, efficient capture of 129I 2 (t 1/2 = 15.7 x 10 6 years) from the nuclear waste stream can help to reduce the risk of releasing I 2 radionuclide into the environment and mitigate concerns about human health problems. Metal organic frameworks (MOFs) have been reported to be potential I 2 adsorbents: but the effect of water vapor, generally present in the reprocessing off-gas stream is rarely taken into account. Moisturemore » stable porous MOFs, which can selectively adsorb I 2 in presence of water vapor is thus of great interest. Herein, the I 2 adsorption performance of two microporous MOFs is reported in presence of different humidity. I…π phenyl ring interactions are mainly responsible for the adsorption as revealed by single crystal XRD« less

Breakthrough in Xenon Capture and Purification Using Adsorbent-Supported Silver Nanoparticles.

PubMed

Deliere, Ludovic; Coasne, Benoit; Topin, Sylvain; Gréau, Claire; Moulin, Christophe; Farrusseng, David

2016-07-04

Rare gas capture and purification is a major challenge for energy, environment, and health applications. Of utmost importance for the nuclear industry, novel separation processes for Xe are urgently needed for spent nuclear fuel reprocessing and nuclear activity monitoring. The recovered, non-radioactive Xe is also of high economic value for lighting, surgical anesthetic, etc. Here, using adsorption and breakthrough experiments and statistical mechanics molecular simulation, we show the outstanding performance of zeolite-supported silver nanoparticles to capture/separate Xe at low concentrations (0.087-100 ppm). We also establish the efficiency of temperature swing adsorption based on such adsorbents for Xe separation from Kr/Xe mixtures and air streams corresponding to off-gases generated by nuclear reprocessing. This study paves the way for the development of novel, cost-efficient technologies relying on the large selectivity/capacity of adsorbent-supported silver nanoparticles which surpass all materials ever tested. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Vertical distributions of (99)Tc and the (99)Tc/(137)Cs activity ratio in the coastal water off Aomori, Japan.

PubMed

Nakanishi, Takahiro; Zheng, Jian; Aono, Tatsuo; Yamada, Masatoshi; Kusakabe, Masashi

2011-08-01

Using a sector-field ICP-MS the vertical distributions of the (99)Tc concentration and (99)Tc/(137)Cs activity ratio were measured in the coastal waters off Aomori Prefecture, Japan, where a spent-nuclear-fuel reprocessing plant has begun test operation. The (99)Tc concentrations in surface water ranged from 1.8 to 2.4 mBq/m(3), no greater than the estimated background level. Relatively high (99)Tc/(137)Cs activity ratios (10-12 × 10(-4)) would be caused by the inflow of the high-(99)Tc/(137)Cs water mass from the Japan Sea. There is no observable contamination from the reprocessing plant in the investigated area. The (99)Tc concentration and the (99)Tc/(137)Cs activity ratio in water column showed gradual decreases with depth. Our results implied that (99)Tc behaves in a more conservative manner than (137)Cs in marine environments. Copyright © 2011 Elsevier Ltd. All rights reserved.

An evaluation of UO2-CNT composites made by SPS as an accident tolerant nuclear fuel pellet and the feasibility of SPS as an economical fabrication process for the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Cartas, Andrew R.

The innovative and advanced purpose of this study is to understand and establish proper sintering procedures for Spark Plasma Sintering process in order to fabricate high density, high thermal conductivity UO2 -CNT pellets. Mixing quality and chemical reactions have been investigated by field emission scanning electron microscopy (FESEM), wavelength dispersive spectroscopy (WDS), and X-ray diffraction (XRD). The effect of various types of CNTs on the mixing and sintering quality of UO2-CNT pellets with SPS processing have been examined. The Archimedes Immersion Method, laser flash method, and FE-SEM will be used to investigate the density, thermal conductivity, grain size, pinning effects, and CNT dispersion of fabricated UO2-CNT pellets. Pre-fabricated CNT's were added to UO 2 powder and dispersed via sonication and/or ball milling and then made into composite nuclear pellets. An investigation of the economic impact of SPS on the nuclear fuel cycle for producing pure and composite UO2 fuels was conducted.

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

Carmack, Jon; Hayes, Steven; Walters, L. C.

This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO 2 and UO 2-PuO 2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availabilitymore » are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.« less

Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

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

Renfro, David; Chandler, David; Cook, David

2014-10-30

Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted usingmore » the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.« less

Final Report on Two-Stage Fast Spectrum Fuel Cycle Options

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

Yang, Won Sik; Lin, C. S.; Hader, J. S.

2016-01-30

This report presents the performance characteristics of two “two-stage” fast spectrum fuel cycle options proposed to enhance uranium resource utilization and to reduce nuclear waste generation. One is a two-stage fast spectrum fuel cycle option of continuous recycle of plutonium (Pu) in a fast reactor (FR) and subsequent burning of minor actinides (MAs) in an accelerator-driven system (ADS). The first stage is a sodium-cooled FR fuel cycle starting with low-enriched uranium (LEU) fuel; at the equilibrium cycle, the FR is operated using the recovered Pu and natural uranium without supporting LEU. Pu and uranium (U) are co-extracted from the dischargedmore » fuel and recycled in the first stage, and the recovered MAs are sent to the second stage. The second stage is a sodium-cooled ADS in which MAs are burned in an inert matrix fuel form. The discharged fuel of ADS is reprocessed, and all the recovered heavy metals (HMs) are recycled into the ADS. The other is a two-stage FR/ADS fuel cycle option with MA targets loaded in the FR. The recovered MAs are not directly sent to ADS, but partially incinerated in the FR in order to reduce the amount of MAs to be sent to the ADS. This is a heterogeneous recycling option of transuranic (TRU) elements« less

Cell module and fuel conditioner

NASA Technical Reports Server (NTRS)

Hoover, D. Q., Jr.

1980-01-01

The computer code for the detailed analytical model of the MK-2 stacks is described. An ERC proprietary matrix is incorporated in the stacks. The mechanical behavior of the stack during thermal cycles under compression was determined. A 5 cell stack of the MK-2 design was fabricated and tested. Designs for the next three stacks were selected and component fabrication initiated. A 3 cell stack which verified the use of wet assembly and a new acid fill procedure were fabricated and tested. Components for the 2 kW test facility were received or fabricated and construction of the facility is underway. The definition of fuel and water is used in a study of the fuel conditioning subsystem. Kinetic data on several catalysts, both crushed and pellets, was obtained in the differential reactor. A preliminary definition of the equipment requirements for treating tap and recovered water was developed.

Process to separate transuranic elements from nuclear waste

DOEpatents

Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

1989-03-21

A process is described for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR). 2 figs.

Method for dissolving plutonium oxide with HI and separating plutonium

DOEpatents

Vondra, Benedict L.; Tallent, Othar K.; Mailen, James C.

1979-01-01

PuO.sub.2 -containing solids, particularly residues from incomplete HNO.sub.3 dissolution of irradiated nuclear fuels, are dissolved in aqueous HI. The resulting solution is evaporated to dryness and the solids are dissolved in HNO.sub.3 for further chemical reprocessing. Alternatively, the HI solution containing dissolved Pu values, can be contacted with a cation exchange resin causing the Pu values to load the resin. The Pu values are selectively eluted from the resin with more concentrated HI.

Process to separate transuranic elements from nuclear waste

DOEpatents

Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

1988-07-12

A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR). 2 figs.

Catalytic Reduction of Hexavalent Uranium by Formic Acid; RIDUZIONE CATALITICA DELL'URANIO ESAVALENTE MEDIANTE ACIDO FORMICO

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

Cogliati, G.; Lanz, R.; Lepscky, C.

1963-10-01

S>The catalytic reduction of U(VI) to U(IV) by means of formic acid has been studied, considering particularly the uranyl nltrate solutions, This process will be applied in the urania--thoria mixed fuel reprocessing plant, (PCUT). Various catalysts have been tested and the influence of formic acid concentration, temperature and catalyst concentration on the reaction rate have been determined. A possible reduction mechanism coherent with Ihe experimental data is discussed. (auth)

Transuranic inventory reduction in repository by partitioning and transmutation

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

Kang, C.H.; Kazimi, M.S.

1992-01-01

The promise of a new reprocessing technology and the issuance of Environmental Protection Agency (EPA) and U.S. Nuclear Regulatory Commission regulations concerning a geologic repository rekindle the interest in partitioning and transmutation of transuranic (TRU) elements from discharged reactor fuel as a high level waste management option. This paper investigates the TRU repository inventory reduction capability of the proposed advanced liquid metal reactors (ALMRs) and integral fast reactors (IFRs) as well as the plutonium recycled light water reactors (LWRs).

Measurement system for alpha emitters in solution

NASA Astrophysics Data System (ADS)

Robert, A.; Sella, C.; Heindl, R.

1984-08-01

The measurement of alpha emitter concentrations in solution corresponds to a need felt in particular by laboratories working on actinides and in the spent fuel reprocessing industry. The instrument present here allows this measurement continuously by the use of a new scintillator that is insensitive to corrosive liquids. The extreme thinness of the scintillator guarantees good detection selectivity of alpha particles in the presence of beta and gamma emissions. Examples of uranium-233, plutonium-239 and americium-241 concentration measurements are presented.

Process to separate transuranic elements from nuclear waste

DOEpatents

Johnson, Terry R.; Ackerman, John P.; Tomczuk, Zygmunt; Fischer, Donald F.

1989-01-01

A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR).

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

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.

Status of the atomized uranium silicide fuel development at KAERI

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

Kim, C.K.; Kim, K.H.; Park, H.D.

1997-08-01

While developing KMRR fuel fabrication technology an atomizing technique has been applied in order to eliminate the difficulties relating to the tough property of U{sub 3}Si and to take advantage of the rapid solidification effect of atomization. The comparison between the conventionally comminuted powder dispersion fuel and the atomized powder dispersion fuel has been made. As the result, the processes, uranium silicide powdering and heat treatment for U{sub 3}Si transformation, become simplified. The workability, the thermal conductivity and the thermal compatibility of fuel meat have been investigated and found to be improved due to the spherical shape of atomized powder.more » In this presentation the overall developments of atomized U{sub 3}Si dispersion fuel and the planned activities for applying the atomizing technique to the real fuel fabrication are described.« less

MEMS-based fuel cells with integrated catalytic fuel processor and method thereof

DOEpatents

Jankowski, Alan F [Livermore, CA; Morse, Jeffrey D [Martinez, CA; Upadhye, Ravindra S [Pleasanton, CA; Havstad, Mark A [Davis, CA

2011-08-09

Described herein is a means to incorporate catalytic materials into the fuel flow field structures of MEMS-based fuel cells, which enable catalytic reforming of a hydrocarbon based fuel, such as methane, methanol, or butane. Methods of fabrication are also disclosed.

Initial Microstructure Evaluation of a U3Si2 + W Fuel Pin Fabricated Via Arc Melt Gravity Drop Casting

NASA Astrophysics Data System (ADS)

Hoggan, Rita E.; Harp, Jason M.

2018-02-01

Injection casting has historically been used to fabricate metallic nuclear fuel on a large scale. Casting of intermetallic fuel forms, such as U3Si2, may be an alternative pathway for fabrication of fuel pins to powder metallurgy. To investigate casting on a small scale, arc melt gravity drop casting was employed to cast a one-off pin of U3Si2 for evaluation as a fabrication method for U3Si2 as a light water reactor fuel. The pin was sectioned and examined via optical microscopy and scanning electron microscopy equipped with energy dispersive x-ray spectroscopy (EDS). Image analysis was used to estimate the volume fraction of phase impurities as well as porosity. The primary phase determined by EDS was U3Si2 with U-O and U-Si-W phase impurities. Unusually high levels of tungsten were observed because of accidental tungsten introduction during arc melting. No significant changes in microstructure were observed after annealing a section of the pin at 800°C for 72 h. The average density of the sectioned specimens was 12.4 g/cm3 measured via Archimedes principle immersion density and He gas displacement.

Fabrication Method for Laboratory-Scale High-Performance Membrane Electrode Assemblies for Fuel Cells.

PubMed

Sassin, Megan B; Garsany, Yannick; Gould, Benjamin D; Swider-Lyons, Karen E

2017-01-03

Custom catalyst-coated membranes (CCMs) and membrane electrode assemblies (MEAs) are necessary for the evaluation of advanced electrocatalysts, gas diffusion media (GDM), ionomers, polymer electrolyte membranes (PEMs), and electrode structures designed for use in next-generation fuel cells, electrolyzers, or flow batteries. This Feature provides a reliable and reproducible fabrication protocol for laboratory scale (10 cm 2 ) fuel cells based on ultrasonic spray deposition of a standard Pt/carbon electrocatalyst directly onto a perfluorosulfonic acid PEM.

Effects of the shape of the foil corners on the irradiation performance of U10Mo alloy based monolithic mini-plates

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

Ozaltun, Hakan; Medvedev, Pavel G

2015-06-01

Monolithic plate-type fuel is a fuel form being developed for high performance research and test reactors to minimize the use of enriched material. These fuel elements are comprised of a high density, low enrichment, U-Mo alloy based fuel foil, sandwiched between Zirconium liners and encapsulated in Aluminum cladding. The use of a high density fuel in a foil form presents a number of fabrication and operational concerns, such as: foil centering, flatness of the foil, fuel thickness variation, geometrical tilting, foil corner shape etc. To benchmark this new design, effects of various geometrical and operational variables on irradiation performance havemore » been evaluated. As a part of these series of sensitivity studies, the shape of the foil corners were studied. To understand the effects of the corner shapes of the foil on thermo-mechanical performance of the plates, a behavioral model was developed for a selected plate from RERTR-12 experiments (Plate L1P785). Both fabrication and irradiation processes were simulated. Once the thermo-mechanical behavior the plate is understood for the nominal case, the simulations were repeated for two additional corner shapes to observe the changes in temperature, displacement and stress-strain fields. The results from the fabrication simulations indicated that the foil corners do not alter the post-fabrication stress-strain magnitudes. Furthermore, the irradiation simulations revealed that post-fabrication stresses of the foil would be relieved very quickly in operation. While, foils with chamfered and filleted corners yielded stresses with comparable magnitudes, they are slightly lower in magnitudes, and provided a more favorable mechanical response compared with the foil with sharp corners.« less

Binder Jetting: A Novel Solid Oxide Fuel-Cell Fabrication Process and Evaluation

NASA Astrophysics Data System (ADS)

Manogharan, Guha; Kioko, Meshack; Linkous, Clovis

2015-03-01

With an ever-growing concern to find a more efficient and less polluting means of producing electricity, fuel cells have constantly been of great interest. Fuel cells electrochemically convert chemical energy directly into electricity and heat without resorting to combustion/mechanical cycling. This article studies the solid oxide fuel cell (SOFC), which is a high-temperature (100°C to 1000°C) ceramic cell made from all solid-state components and can operate under a wide range of fuel sources such as hydrogen, methanol, gasoline, diesel, and gasified coal. Traditionally, SOFCs are fabricated using processes such as tape casting, calendaring, extrusion, and warm pressing for substrate support, followed by screen printing, slurry coating, spray techniques, vapor deposition, and sputter techniques, which have limited control in substrate microstructure. In this article, the feasibility of engineering the porosity and configuration of an SOFC via an additive manufacturing (AM) method known as binder jet printing was explored. The anode, cathode and oxygen ion-conducting electrolyte layers were fabricated through AM sequentially as a complete fuel cell unit. The cell performance was measured in two modes: (I) as an electrolytic oxygen pump and (II) as a galvanic electricity generator using hydrogen gas as the fuel. An analysis on influence of porosity was performed through SEM studies and permeability testing. An additional study on fuel cell material composition was conducted to verify the effects of binder jetting through SEM-EDS. Electrical discharge of the AM fabricated SOFC and nonlinearity of permeability tests show that, with additional work, the porosity of the cell can be modified for optimal performance at operating flow and temperature conditions.

10 CFR 75.4 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

...); (3) A fuel fabrication plant; (4) An enrichment plant or isotope separation plant for the separation..., irradiated fuel element chopping machines, and hot cells. Nuclear fuel cycle-related research and development...

10 CFR 75.4 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

...); (3) A fuel fabrication plant; (4) An enrichment plant or isotope separation plant for the separation..., irradiated fuel element chopping machines, and hot cells. Nuclear fuel cycle-related research and development...

U-10Mo Baseline Fuel Fabrication Process Description

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

Hubbard, Lance R.; Arendt, Christina L.; Dye, Daniel F.

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

Development of Impregnated Agglomerate Pelletization (IAP) process for fabrication of (Th,U)O 2 mixed oxide pellets

NASA Astrophysics Data System (ADS)

Khot, P. M.; Nehete, Y. G.; Fulzele, A. K.; Baghra, Chetan; Mishra, A. K.; Afzal, Mohd.; Panakkal, J. P.; Kamath, H. S.

2012-01-01

Impregnated Agglomerate Pelletization (IAP) technique has been developed at Advanced Fuel Fabrication Facility (AFFF), BARC, Tarapur, for manufacturing (Th, 233U)O 2 mixed oxide fuel pellets, which are remotely fabricated in hot cell or shielded glove box facilities to reduce man-rem problem associated with 232U daughter radionuclides. This technique is being investigated to fabricate the fuel for Indian Advanced Heavy Water Reactor (AHWR). In the IAP process, ThO 2 is converted to free flowing spheroids by powder extrusion route in an unshielded facility which are then coated with uranyl nitrate solution in a shielded facility. The dried coated agglomerate is finally compacted and then sintered in oxidizing/reducing atmosphere to obtain high density (Th,U)O 2 pellets. In this study, fabrication of (Th,U)O 2 mixed oxide pellets containing 3-5 wt.% UO 2 was carried out by IAP process. The pellets obtained were characterized using optical microscopy, XRD and alpha autoradiography. The results obtained were compared with the results for the pellets fabricated by other routes such as Coated Agglomerate Pelletization (CAP) and Powder Oxide Pelletization (POP) route.

Design and testing of digitally manufactured paraffin Acrylonitrile-butadiene-styrene hybrid rocket motors

NASA Astrophysics Data System (ADS)

McCulley, Jonathan M.

This research investigates the application of additive manufacturing techniques for fabricating hybrid rocket fuel grains composed of porous Acrylonitrile-butadiene-styrene impregnated with paraffin wax. The digitally manufactured ABS substrate provides mechanical support for the paraffin fuel material and serves as an additional fuel component. The embedded paraffin provides an enhanced fuel regression rate while having no detrimental effect on the thermodynamic burn properties of the fuel grain. Multiple fuel grains with various ABS-to-Paraffin mass ratios were fabricated and burned with nitrous oxide. Analytical predictions for end-to-end motor performance and fuel regression are compared against static test results. Baseline fuel grain regression calculations use an enthalpy balance energy analysis with the material and thermodynamic properties based on the mean paraffin/ABS mass fractions within the fuel grain. In support of these analytical comparisons, a novel method for propagating the fuel port burn surface was developed. In this modeling approach the fuel cross section grid is modeled as an image with white pixels representing the fuel and black pixels representing empty or burned grid cells.

Prototype Demonstration of Gamma- Blind Tensioned Metastable Fluid Neutron/Multiplicity/Alpha Detector – Real Time Methods for Advanced Fuel Cycle Applications

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

McDeavitt, Sean M.

The content of this report summarizes a multi-year effort to develop prototype detection equipment using the Tensioned Metastable Fluid Detector (TMFD) technology developed by Taleyarkhan [1]. The context of this development effort was to create new methods for evaluating and developing advanced methods for safeguarding nuclear materials along with instrumentation in various stages of the fuel cycle, especially in material balance areas (MBAs) and during reprocessing of used nuclear fuel. One of the challenges related to the implementation of any type of MBA and/or reprocessing technology (e.g., PUREX or UREX) is the real-time quantification and control of the transuranic (TRU)more » isotopes as they move through the process. Monitoring of higher actinides from their neutron emission (including multiplicity) and alpha signatures during transit in MBAs and in aqueous separations is a critical research area. By providing on-line real-time materials accountability, diversion of the materials becomes much more difficult. The Tensioned Metastable Fluid Detector (TMFD) is a transformational technology that is uniquely capable of both alpha and neutron spectroscopy while being “blind” to the intense gamma field that typically accompanies used fuel – simultaneously with the ability to provide multiplicity information as well [1-3]. The TMFD technology was proven (lab-scale) as part of a 2008 NERI-C program [1-7]. The bulk of this report describes the advancements and demonstrations made in TMFD technology. One final point to present before turning to the TMFD demonstrations is the context for discussing real-time monitoring of SNM. It is useful to review the spectrum of isotopes generated within nuclear fuel during reactor operations. Used nuclear fuel (UNF) from a light water reactor (LWR) contains fission products as well as TRU elements formed through neutron absorption/decay chains. The majority of the fission products are gamma and beta emitters and they represent the more significant hazards from a radiation protection standpoint. However, alpha and neutron emitting uranium and TRU elements represent the more significant safeguards and security concerns. Table 1.1 presents a representative PWR inventory of the uranium and actinide isotopes present in a used fuel assembly. The uranium and actinide isotopes (chiefly the Pu, Am and Cm elements) are all emitters of alpha particles and some of them release significant quantities of neutrons through spontaneous fissions« less

Development of Metallic Fuels for Actinide Transmutation

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

Hayes, Steven Lowe; Fielding, Randall Sidney; Benson, Michael Timothy

Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimizedmore » for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10 and 60%. In general, the performance of all of these substantially disparate metallic fuel alloys has been observed to be excellent, and their irradiation behaviors are generally consistent with historic norms for metallic fuels without minor actinide additions and having lower Pu or Zr contents. Future work is being undertaken with a view toward increasing the burnup potential of metallic fuels even more. Design innovations under investigation include: 1) lowering the fuel smear density in order to accommodate more swelling, 2) annular fuel geometry to eliminate the need for a sodium bond, 3) minor alloy additions to stabilize lanthanide fission products inside the fuel and prevent their transport to the cladding where they can participate in fuel-cladding chemical interaction (FCCI), and 4) coatings/liners on the cladding inner surface to mitigate FCCI and enable higher temperature operation. This paper will present the current state of development of metallic fuels for actinide transmutation in the US. Highlights will include recent results from metallic fuel casting experiments, experiments to identify alloy additions to immobilize lanthanide fission products, and postirradiation examinations of annular metallic fuels at low burnup.« less

Organic fuel cell methods and apparatus

NASA Technical Reports Server (NTRS)

Narayanan, Sekharipuram R. (Inventor); Frank, Harvey A. (Inventor); Halpert, Gerald (Inventor); Surampudi, Subbarao (Inventor); Prakash, G. K. Surya (Inventor); Vamos, Eugene (Inventor); Olah, George A. (Inventor)

2001-01-01

A liquid organic fuel cell is provided which employs a solid electrolyte membrane. An organic fuel, such as a methanol/water mixture, is circulated past an anode of a cell while oxygen or air is circulated past a cathode of the cell. The cell solid electrolyte membrane is preferably fabricated from Nafion.TM.. Additionally, a method for improving the performance of carbon electrode structures for use in organic fuel cells is provided wherein a high surface-area carbon particle/Teflon.TM.-binder structure is immersed within a Nafion.TM./methanol bath to impregnate the electrode with Nafion.TM.. A method for fabricating an anode for use in a organic fuel cell is described wherein metal alloys are deposited onto the electrode in an electro-deposition solution containing perfluorooctanesulfonic acid. A fuel additive containing perfluorooctanesulfonic acid for use with fuel cells employing a sulfuric acid electrolyte is also disclosed. New organic fuels, namely, trimethoxymethane, dimethoxymethane, and trioxane are also described for use with either conventional or improved fuel cells.

Organic fuel cell methods and apparatus

NASA Technical Reports Server (NTRS)

Surampudi, Subbarao (Inventor); Narayanan, Sekharipuram R. (Inventor); Vamos, Eugene (Inventor); Halpert, Gerald (Inventor); Olah, George A. (Inventor); Frank, Harvey A. (Inventor); Prakash, G. K. Surya (Inventor)

2004-01-01

A liquid organic, fuel cell is provided which employs a solid electrolyte membrane. An organic fuel, such as a methanol/water mixture, is circulated past an anode of a cell while oxygen or air is circulated past a cathode of the cell. The cell solid electrolyte membrane is preferably fabricated from Nafion.TM.. Additionally, a method for improving the performance of carbon electrode structures for use in organic fuel cells is provided wherein a high surface-area carbon particle/Teflon.TM.-binder structure is immersed within a Nafion.TM./methanol bath to impregnate the electrode with Nafion.TM.. A method for fabricating an anode for use in a organic fuel cell is described wherein metal alloys are deposited onto the electrode in an electro-deposition solution containing perfluorooctanesulfonic acid. A fuel additive containing perfluorooctanesulfonic acid for use with fuel cells employing a sulfuric acid electrolyte is also disclosed. New organic fuels, namely, trimethoxymethane, dimethoxymethane, and trioxane are also described for use with either conventional or improved fuel cells.

Organic fuel cell methods and apparatus

NASA Technical Reports Server (NTRS)

Frank, Harvey A. (Inventor); Halpert, Gerald (Inventor); Surampudi, Subbarao (Inventor); Olah, George A. (Inventor); Vamos, Eugene (Inventor); Narayanan, Sekharipuram R. (Inventor); Prakash, G. K. Surya (Inventor)

2008-01-01

A liquid organic, fuel cell is provided which employs a solid electrolyte membrane. An organic fuel, such as a methanol/water mixture, is circulated past an anode of a cell while oxygen or air is circulated past a cathode of the cell. The cell solid electrolyte membrane is preferably fabricated from Nafion.TM.. Additionally, a method for improving the performance of carbon electrode structures for use in organic fuel cells is provided wherein a high surface-area carbon particle/Teflon.TM.-binder structure is immersed within a Nafion.TM./methanol bath to impregnate the electrode with Nafion.TM.. A method for fabricating an anode for use in a organic fuel cell is described wherein metal alloys are deposited onto the electrode in an electro-deposition solution containing perfluorooctanesulfonic acid. A fuel additive containing perfluorooctanesulfonic acid for use with fuel cells employing a sulfuric acid electrolyte is also disclosed. New organic fuels, namely, trimethoxymethane, dimethoxymethane, and trioxane are also described for use with either conventional or improved fuel cells.

Aqueous liquid feed organic fuel cell using solid polymer electrolyte membrane

NASA Technical Reports Server (NTRS)

Olah, George A. (Inventor); Surampudi, Subbarao (Inventor); Vamos, Eugene (Inventor); Halpert, Gerald (Inventor); Narayanan, Sekharipuram R. (Inventor); Frank, Harvey A. (Inventor); Prakash, G. K. Surya (Inventor)

1997-01-01

A liquid organic fuel cell is provided which employs a solid electrolyte membrane. An organic fuel, such as a methanol/water mixture, is circulated past an anode of a cell while oxygen or air is circulated past a cathode of the cell. The cell solid electrolyte membrane is preferably fabricated from Nafion.TM.. Additionally, a method for improving the performance of carbon electrode structures for use in organic fuel cells is provided wherein a high surface-area carbon particle/Teflon.TM.-binder structure is immersed within a Nafion.TM./methanol bath to impregnate the electrode with Nafion.TM.. A method for fabricating an anode for use in a organic fuel cell is described wherein metal alloys are deposited onto the electrode in an electro-deposition solution containing perfluorooctanesulfonic acid. A fuel additive containing perfluorooctanesulfonic acid for use with fuel cells employing a sulfuric acid electrolyte is also disclosed. New organic fuels, namely, trimethoxymethane, dimethoxymethane, and trioxane are also described for use with either conventional or improved fuel cells.

Fuel cell system with sodium borohydride as hydrogen source for unmanned aerial vehicles

NASA Astrophysics Data System (ADS)

Kim, Kyunghwan; Kim, Taegyu; Lee, Kiseong; Kwon, Sejin

In this study, we design and fabricate a fuel cell system for application as a power source in unmanned aerial vehicles (UAVs). The fuel cell system consists of a fuel cell stack, hydrogen generator, and hybrid power management system. PEMFC stack with an output power of 100 W is prepared and tested to decide the efficient operating conditions; the stack must be operated in the dead-end mode with purge in order to ensure prolonged stack performance. A hydrogen generator is fabricated to supply gaseous hydrogen to the stack. Sodium borohydride (NaBH 4) is used as the hydrogen source in the present study. Co/Al 2O 3 catalyst is prepared for the hydrolysis of the alkaline NaBH 4 solution at room temperature. The fabricated Co catalyst is comparable to the Ru catalyst. The UAV consumes more power in the takeoff mode than in the cruising mode. A hybrid power management system using an auxiliary battery is developed and evaluated for efficient energy management. Hybrid power from both the fuel cell and battery powers takeoff and turning flight operations, while the fuel cell supplies steady power during the cruising flight. The capabilities of the fuel-cell UAVs for long endurance flights are validated by successful flight tests.

Laboratory Scale Coal And Biomass To Drop-In Fuels (CBDF) Production And Assessment

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

Lux, Kenneth; Imam, Tahmina; Chevanan, Nehru

This Final Technical Report describes the work and accomplishments of the project entitled, “Laboratory Scale Coal and Biomass to Drop-In Fuels (CBDF) Production and Assessment.” The main objective of the project was to fabricate and test a lab-scale liquid-fuel production system using coal containing different percentages of biomass such as corn stover and switchgrass at a rate of 2 liters per day. The system utilizes the patented Altex fuel-production technology, which incorporates advanced catalysts developed by Pennsylvania State University. The system was designed, fabricated, tested, and assessed for economic and environmental feasibility relative to competing technologies.

Solid Polymer Electrolyte Fuel Cell Technology Program

NASA Technical Reports Server (NTRS)

1980-01-01

Work is reported on phase 5 of the Solid Polymer Electrolyte (SPE) Fuel Cell Technology Development program. The SPE fuel cell life and performance was established at temperatures, pressures, and current densities significantly higher than those previously demonstrated in sub-scale hardware. Operation of single-cell Buildup No. 1 to establish life capabilities of the full-scale hardware was continued. A multi-cell full-scale unit (Buildup No. 2) was designed, fabricated, and test evaluated laying the groundwork for the construction of a reactor stack. A reactor stack was then designed, fabricated, and successfully test-evaluated to demonstrate the readiness of SPE fuel cell technology for future space applications.

Fuel cycle cost uncertainty from nuclear fuel cycle comparison

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

Li, J.; McNelis, D.; Yim, M.S.

2013-07-01

This paper examined the uncertainty in fuel cycle cost (FCC) calculation by considering both model and parameter uncertainty. Four different fuel cycle options were compared in the analysis including the once-through cycle (OT), the DUPIC cycle, the MOX cycle and a closed fuel cycle with fast reactors (FR). The model uncertainty was addressed by using three different FCC modeling approaches with and without the time value of money consideration. The relative ratios of FCC in comparison to OT did not change much by using different modeling approaches. This observation was consistent with the results of the sensitivity study for themore » discount rate. Two different sets of data with uncertainty range of unit costs were used to address the parameter uncertainty of the FCC calculation. The sensitivity study showed that the dominating contributor to the total variance of FCC is the uranium price. In general, the FCC of OT was found to be the lowest followed by FR, MOX, and DUPIC. But depending on the uranium price, the FR cycle was found to have lower FCC over OT. The reprocessing cost was also found to have a major impact on FCC.« less

A physical and economic model of the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Schneider, Erich Alfred

A model of the nuclear fuel cycle that is suitable for use in strategic planning and economic forecasting is presented. The model, to be made available as a stand-alone software package, requires only a small set of fuel cycle and reactor specific input parameters. Critical design criteria include ease of use by nonspecialists, suppression of errors to within a range dictated by unit cost uncertainties, and limitation of runtime to under one minute on a typical desktop computer. Collision probability approximations to the neutron transport equation that lead to a computationally efficient decoupling of the spatial and energy variables are presented and implemented. The energy dependent flux, governed by coupled integral equations, is treated by multigroup or continuous thermalization methods. The model's output includes a comprehensive nuclear materials flowchart that begins with ore requirements, calculates the buildup of 24 actinides as well as fission products, and concludes with spent fuel or reprocessed material composition. The costs, direct and hidden, of the fuel cycle under study are also computed. In addition to direct disposal and plutonium recycling strategies in current use, the model addresses hypothetical cycles. These include cycles chosen for minor actinide burning and for their low weapons-usable content.

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

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

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

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

Mass tracking and material accounting in the integral fast reactor (IFR)

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

Orechwa, Y.; Adams, C.H.; White, A.M.

1991-01-01

This paper reports on the Integral Fast Reactor (IFR) which is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory. There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure with compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstratedmore » in the facilities at ANL-West, utilizing Experimental Breeder Reactor II and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations.« less

Field test of short-notice random inspections for inventory-change verification at a low-enriched-uranium fuel-fabrication plant: Preliminary summary

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

Fishbone, L.G.; Moussalli, G.; Naegele, G.

1994-04-01

An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. This report details a six-month field test of the feasibility of such SNRIs which took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division. Westinghouse personnel made daily declarations aboutmore » both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ``mailbox``. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. Items from both strata were verified during the SNRIs by means of nondestructive assay equipment. The field test demonstrated the feasibility and practicality of key elements of the SNRI approach for a large LEU fuel fabrication plant.« less

Development of a new casting method to fabricate U–Zr alloy containing minor actinides

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

Jong Hwan Kim; Hoon Song; Hyung Tae Kim

2014-01-01

Metal fuel slugs of U–Zr alloys for a sodium-cooled fast reactor (SFR) have conventionally been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents, such as Am, are problematic in a conventional injection casting method. As an alternative fabrication method, low pressure gravity casting has been developed. Casting soundness, microstructural characteristics, alloying composition, density, and fuel losses were evaluated for the following as-cast fuel slugs: U–10 wt% Zr, U–10 wt% Zr–5 wt% RE, and U–10 wt% Zr–5 wt% RE–5 wt% Mn. The U and Zr contents were uniform throughout the matrix, and impurities such as oxyen,more » carbon, and nitrogen satisfied the specification of total impurities less than 2,000 ppm. The appearance of the fuel slugs was generally sound, and the internal integrity was shown to be satisfactory based on gamma-ray radiography. In a volatile surrogate casting test, the U–Zr–RE–Mn fuel slug showed that nearly all of the manganese was retained when casting was done under an inert atmosphere.« less

Microfabrication of microchannels for fuel cell plates.

PubMed

Jang, Ho Su; Park, Dong Sam

2010-01-01

Portable electronic devices such as notebook computers, PDAs, cellular phones, etc., are being widely used, and they increasingly need cheap, efficient, and lightweight power sources. Fuel cells have been proposed as possible power sources to address issues that involve energy production and the environment. In particular, a small type of fuel-cell system is known to be suitable for portable electronic devices. The development of micro fuel cell systems can be achieved by the application of microchannel technology. In this study, the conventional method of chemical etching and the mechanical machining method of micro end milling were used for the microfabrication of microchannel for fuel cell separators. The two methods were compared in terms of their performance in the fabrication with regards to dimensional errors, flatness, straightness, and surface roughness. Following microchannel fabrication, the powder blasting technique is introduced to improve the coating performance of the catalyst on the surface of the microchannel. Experimental results show that end milling can remarkably increase the fabrication performance and that surface treatment by powder blasting can improve the performance of catalyst coating.

Microfabrication of Microchannels for Fuel Cell Plates

PubMed Central

Jang, Ho Su; Park, Dong Sam

2010-01-01

Portable electronic devices such as notebook computers, PDAs, cellular phones, etc., are being widely used, and they increasingly need cheap, efficient, and lightweight power sources. Fuel cells have been proposed as possible power sources to address issues that involve energy production and the environment. In particular, a small type of fuel-cell system is known to be suitable for portable electronic devices. The development of micro fuel cell systems can be achieved by the application of microchannel technology. In this study, the conventional method of chemical etching and the mechanical machining method of micro end milling were used for the microfabrication of microchannel for fuel cell separators. The two methods were compared in terms of their performance in the fabrication with regards to dimensional errors, flatness, straightness, and surface roughness. Following microchannel fabrication, the powder blasting technique is introduced to improve the coating performance of the catalyst on the surface of the microchannel. Experimental results show that end milling can remarkably increase the fabrication performance and that surface treatment by powder blasting can improve the performance of catalyst coating. PMID:22315533

Monitoring iodine-129 in air and milk samples collected near the Hanford Site: an investigation of historical iodine monitoring data.

PubMed

Fritz, Brad G; Patton, Gregory W

2006-01-01

While other research has reported on the concentrations of (129)I in the environment surrounding active nuclear fuel reprocessing facilities, there is a shortage of information regarding how the concentrations change once facilities close. At the Hanford Site, the Plutonium-Uranium Extraction (PUREX) chemical separation plant was operating between 1983 and 1990, during which time (129)I concentrations in air and milk were measured. After the cessation of chemical processing, plant emissions decreased 2.5 orders of magnitude over an 8-year period. An evaluation of (129)I and (127)I concentration data in air and milk spanning the PUREX operation and post-closure period was conducted to compare the changes in environmental levels. Measured concentrations over the monitoring period were below the levels that could result in a potential annual human dose greater than 1 mSv. There was a measurable difference in the measured air concentrations of (129)I at different distances from the source, indicating a distinct Hanford fingerprint. Correlations between stack emissions of (129)I and concentrations in air and milk indicate that atmospheric emissions were the major source of (129)I measured in environmental samples. The measured concentrations during PUREX operations were similar to observations made around a fuel reprocessing plant in Germany. After the PUREX Plant stopped operating, (129)I concentration measurements made upwind of Hanford were similar to the results from Seville, Spain.

Brief overview of the long-lived radionuclide separation processes developed in france in connection with the spin program

NASA Astrophysics Data System (ADS)

Madic, Charles; Bourges, Jacques; Dozol, Jean-François

1995-09-01

To reduce the long-term potential hazards associated with the management of nuclear wastes generated by nuclear fuel reprocessing, one alternative is the transmutation of long-lived radionuclides into short-lived radionuclides by nuclear means (P & T strategy). In this context, according to the law passed by the French Parliament on 30 December 1991, the CEA launched the SPIN program for the design of long-lived radionuclide separation and nuclear incineration processes. The research in progress to define separation processes focused mainly on the minor actinides (neptunium, americium and curium) and some fission products, like cesium and technetium. To separate these long-lived radionuclides, two strategies were developed. The first involves research on new operating conditions for improving the PUREX fuel reprocessing technology. This approach concerns the elements neptunium and technetium (iodine and zirconium can also be considered). The second strategy involves the design of new processes; DIAMEX for the co-extraction of minor actinides from the high-level liquid waste leaving the PUREX process, An(III)/Ln(III) separation using tripyridyltriazine derivatives or picolinamide extracting agents; SESAME for the selective separation of americium after its oxidation to Am(IV) or Am(VI) in the presence of a heteropolytungstate ligand, and Cs extraction using a new class of extracting agents, calixarenes, which exhibit exceptional Cs separation properties, especially in the presence of sodium ion. This lecture focuses on the latest achievements in these research areas.

Composition for detecting uranyl

DOEpatents

Baylor, L.C.; Stephens, S.M.

1994-01-01

The present invention relates to an indicator composition for use in spectrophotometric detection of a substance in a solution, and a method for making the composition. Useful indicators are sensitive to the particular substance being measured, but are unaffected by the fluid and other chemical species that may be present in the fluid. Optical indicators are used to measure the uranium concentration of process solutions in facilities for extracting uranium from ores, production of nuclear fuels, and reprocessing of irradiated fuels. The composition comprises an organohalide covalently bonded to an indicator for the substance, in such a manner that the product is itself an indicator that provides increased spectral resolution for detecting the substance. The indicator is preferably arsenazo III and the organohalide is preferably cyanuric chloride. These form a composition that is ideally suited for detecting uranyl.

Applications of plasma core reactors to terrestrial energy systems

NASA Technical Reports Server (NTRS)

Latham, T. S.; Biancardi, F. R.; Rodgers, R. J.

1974-01-01

Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

77 FR 70193 - Shaw Areva MOX Services (Mixed Oxide Fuel Fabrication Facility); Notice of Atomic Safety and...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-23

... MOX Services (Mixed Oxide Fuel Fabrication Facility); Notice of Atomic Safety and Licensing Board Reconstitution Pursuant to 10 CFR 2.313(c) and 2.321(b), the Atomic Safety and Licensing Board (Board) in the... Rockville, Maryland this 16th day of November 2012. E. Roy Hawkens, Chief Administrative Judge, Atomic...

PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY

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

S. T. Khericha

2007-04-01

The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed tomore » achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.« less

Microstructural characterization of a thin film ZrN diffusion barrier in an As-fabricated U-7Mo/Al matrix dispersion fuel plate

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Perez, Emmanuel; Wiencek, Tom; Leenaers, Ann; Van den Berghe, Sven

2015-03-01

The United States High Performance Research Reactor Fuel Development program is developing low enriched uranium fuels for application in research and test reactors. One concept utilizes U-7 wt.% Mo (U-7Mo) fuel particles dispersed in Al matrix, where the fuel particles are coated with a 1 μm-thick ZrN coating. The ZrN serves as a diffusion barrier to eliminate a deleterious reaction that can occur between U-7Mo and Al when a dispersion fuel is irradiated under aggressive reactor conditions. To investigate the final microstructure of a physically-vapor-deposited ZrN coating in a dispersion fuel plate after it was fabricated using a rolling process, characterization samples were taken from a fuel plate that was fabricated at 500 °C using ZrN-coated U-7Mo particles, Al matrix and AA6061 cladding. Scanning electron and transmission electron microscopy analysis were performed. Data from these analyses will be used to support future microstructural examinations of irradiated fuel plates, in terms of understanding the effects of irradiation on the ZrN microstructure, and to determine the role of diffusion barrier microstructure in eliminating fuel/matrix interactions during irradiation. The as-fabricated coating was determined to be cubic-ZrN (cF8) phase. It exhibited a columnar microstructure comprised of nanometer-sized grains and a region of relatively high porosity, mainly near the Al matrix. Small impurity-containing phases were observed at the U-7Mo/ZrN interface, and no interaction zone was observed at the ZrN/Al interface. The bonding between the U-7Mo and ZrN appeared to be mechanical in nature. A relatively high level of oxygen was observed in the ZrN coating, extending from the Al matrix in the ZrN coating in decreasing concentration. The above microstructural characteristics are discussed in terms of what may be most optimal for a diffusion barrier in a dispersion fuel plate application.

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

Analysis and Development of A Robust Fuel for Gas-Cooled Fast Reactors

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

Knight, Travis W.

2010-01-31

The focus of this effort was on the development of an advanced fuel for gas-cooled fast reactor (GFR) applications. This composite design is based on carbide fuel kernels dispersed in a ZrC matrix. The choice of ZrC is based on its high temperature properties and good thermal conductivity and improved retention of fission products to temperatures beyond that of traditional SiC based coated particle fuels. A key component of this study was the development and understanding of advanced fabrication techniques for GFR fuels that have potential to reduce minor actinide (MA) losses during fabrication owing to their higher vapor pressuresmore » and greater volatility. The major accomplishments of this work were the study of combustion synthesis methods for fabrication of the ZrC matrix, fabrication of high density UC electrodes for use in the rotating electrode process, production of UC particles by rotating electrode method, integration of UC kernels in the ZrC matrix, and the full characterization of each component. Major accomplishments in the near-term have been the greater characterization of the UC kernels produced by the rotating electrode method and their condition following the integration in the composite (ZrC matrix) following the short time but high temperature combustion synthesis process. This work has generated four journal publications, one conference proceeding paper, and one additional journal paper submitted for publication (under review). The greater significance of the work can be understood in that it achieved an objective of the DOE Generation IV (GenIV) roadmap for GFR Fuel—namely the demonstration of a composite carbide fuel with 30% volume fuel. This near-term accomplishment is even more significant given the expected or possible time frame for implementation of the GFR in the years 2030 -2050 or beyond.« less

Glucose Fuel Cells with a MicroChannel Fabricated on Flexible Polyimide Film

NASA Astrophysics Data System (ADS)

Sano, Ryohei; Fukushi, Yudai; Sasaki, Tsubasa; Mogi, Hiroshi; Koide, Syohei; Ikoma, Ryuta; Akatsuka, Wataru; Tsujimura, Seiya; Nishioka, Yasushiro

2013-12-01

In this work, a glucose fuel cell was fabricated using microfabrication processes assigned for microelectromechanical systems. The fuel cell was equipped with a microchannel to flow an aqueous solution of glucose. The cell was fabricated on a flexible polyimide substrate, and its porous carbon-coated aluminum (Al) electrodes of 2.8 mm in width and 11 mm in length were formed using photolithography and screen printing techniques. Porous carbon was deposited by screen printing of carbon black ink on the Al electrode surfaces in order to increase the effective electrode surface area and to absorb more enzymes on the electrode surfaces. The microchannel with a depth of 200 μm was fabricated using a hot embossing technique. A maximum power of 0.45 μW at 0.5 V that corresponds to a power density of 1.45 μW/cm2 was realized by introducing a 200 mM concentrated glucose solution at room temperature.

NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

DOEpatents

Stengel, F.G.

1963-12-24

A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

Electrodeposition for Electrochemical Energy Conversion and Storage Devices

NASA Astrophysics Data System (ADS)

Shaigan, Nima

Electrodeposition of metals, alloys, metal oxides, conductive polymers, and their composites plays a pivotal role in fabrication processes of some recently developed electrochemical energy devices, most particularly fuel cells, supercapacitors, and batteries. Unique nanoscale architectures of electrocatalysts for low temperature fuel cells, including proton exchange membrane fuel cell (PEMFC) and direct methanol fuel cell (DMFC), can only be obtained through electrodeposition processes. Promising, cost-effective conductive/protective coatings for stainless steel interconnects used in solid oxide fuel cells (SOFCs) have been achieved employing a variety of electrodeposition techniques. In supercapacitors, anodic deposition of metal oxides, conductive polymers, and their composites is a versatile technique for fabrication of electrodes with distinctive morphology and exceptional specific capacitance. Electrodeposition is also very recently employed for preparation of Sn-based anodes for lithium ion batteries.

Fabrication of TREAT Fuel with Increased Graphite Loading

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

Luther, Erik Paul; Leckie, Rafael M.; Dombrowski, David E.

2014-02-05

As part of the feasibility study exploring the replacement of the HEU fuel core of the TREAT reactor at Idaho National Laboratory with LEU fuel, this study demonstrates that it is possible to increase the graphite content of extruded fuel by reformulation. The extrusion process was use to fabricate the “upgrade” core1 for the TREAT reactor. The graphite content achieved is determined by calculation and has not been measured by any analytical method. In conjunction, a technique, Raman Spectroscopy, has been investigated for measuring the graphite content. This method shows some promise in differentiating between carbon and graphite; however, standardsmore » that would allow the technique to be calibrated to quantify the graphite concentration have yet to be fabricated. Continued research into Raman Spectroscopy is on going. As part of this study, cracking of graphite extrusions due to volatile evolution during heat treatment has been largely eliminated. Continued research to optimize this extrusion method is required.« less