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

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.

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.

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

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.

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

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

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.

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

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

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...

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.

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.

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.

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

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

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.

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

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

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.

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.

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

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.

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

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

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

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.

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

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

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

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.

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

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 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 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.

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

DOE PAGES

Ham, Y.; Kerr, P.; Sitaraman, S.; ...

2016-05-05

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

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

Ham, Y.S.; Kerr, P.; Sitaraman, S.

The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported themore » successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)« less

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

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

Ham, Y.; Kerr, P.; Sitaraman, S.

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

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

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.

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.

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.

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

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

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. ...

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

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

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

The measurement of U(VI) and Np(IV) mass transfer in a single stage centrifugal contactor

NASA Astrophysics Data System (ADS)

May, I.; Birkett, E. J.; Denniss, I. S.; Gaubert, E. T.; Jobson, M.

2000-07-01

BNFL currently operates two reprocessing plants for the conversion of spent nuclear fuel into uranium and plutonium products for fabrication into uranium oxide and mixed uranium and plutonium oxide (MOX) fuels. To safeguard the future commercial viability of this process, BNFL is developing novel single cycle flowsheets that can be operated in conjunction with intensified centrifugal contactors.

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

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

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.

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.

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.

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

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

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

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.

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

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

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

Beatty, Randy L; Harrison, Thomas J

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

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.

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.

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.

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

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

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

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

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

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.

The behaviour of transuranic mixed oxide fuel in a Candu-900 reactor

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

Morreale, A. C.; Ball, M. R.; Novog, D. R.

2012-07-01

The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation and heat loads of spent material. The burning of transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while reducing the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxidemore » fuel, TRUMOX, in a CANDU{sup R}* nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, super-cell calculations were analyzed in DRAGON and full core analysis was executed in the RFSP 2-group diffusion code. A time-average full core model was produced and analyzed for reactor coefficients, reactivity device worth and online fuelling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 27.36 MWd/kg HE. A full TRUMOX fuelled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle. (authors)« less

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.

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.

Gamma-ray mirror technology for NDA of spent fuel

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

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

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

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

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

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.

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

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 and plutonium recovery within the context of the development of nuclear energy in Russia

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

Kushnikov, V.

1996-05-01

The Russian strategy for radioactive waste and plutonium management is based on the concept of the closed fuel cycle that has been adopted in Russia, and, to a great degree, falls under the jurisdiction of the existing Russian nuclear energy structures. From its very beginning, Russian atomic energy policy was based on finding the most effective method of developing the new fuel direction with the maximum possible utilization of the energy potential from the fission of heavy atoms and the achievement of fuel self-sufficiency through the recycling of secondary fuel. Although there can be no doubt about the importance ofmore » economic considerations (for the future), concerns for the safety of the environment are currently of the utmost importance. In this context, spent NPP fuel can be viewed as a waste to be buried only if there is persuasive evidence that such an approach is both economically and environmentally sound. The production of I GW of energy per year is accompanied by the accumulation of up to 800-1000 kg of highly radioactive fission products and approximately 250 kg of plutonium. Currently, spent fuel from the VVER 100 and the RBNK reactors contains approximately 25 tons of plutonium. There is an additional 30 tons of fuel-grade plutonium in the form of purified oxide, separated from spent fuels used in VVER440 reactors and other power production facilities, as well as approximately 100 tons of weapons-grade plutonium from dismantled warheads. The spent fuel accumulates significant amounts of small actinoids - neptunium americium, and curium. Science and technology have not yet found technical solutions for safe and secure burial of non-reprocessed spent fuel with such a broad range of products, which are typically highly radioactive and will continue to pose a threat for hundreds of thousands of years.« 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

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

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.

Preparation of the Second Shipment of Spent Nuclear Fuel from the Ustav Jaderneho Vyzkumu Rez (UJV Rez), a.s., Czech Republic to the Russian Federation for Reprocessing - 13478

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

Trtilek, Radek; Podlaha, Josef

After more than 50 years of operation of the LVR-15 research reactor operated by the UJV Rez, a. s. (formerly Nuclear Research Institute - NRI), a large amount of the spent nuclear fuel (SNF) of Russian origin has been accumulated. In 2005 UJV Rez, a. s. jointed the Russian Research Reactor Fuel Return (RRRFR) program under the United States (US) - Russian Global Threat Reduction Initiative (GTRI) and started the process of SNF shipment from the LVR-15 research reactor back to the Russian Federation (RF). In 2007 the first shipment of SNF was realized. In 2011, preparation of the secondmore » shipment of spent fuel from the Czech Republic started. The experience obtained from the first shipment will be widely used, but some differences must be taken into the account. The second shipment will be realized in 2013 and will conclude the return transport of all, both fresh and spent, high-enriched nuclear fuel from the Czech Republic to the Russian Federation. After the shipment is completed, there will be only low-enriched nuclear fuel on the territory of the Czech Republic, containing maximum of 20% of U-235, which is the conventionally recognized limit between the low- and high-enriched nuclear materials. The experience (technical, organizational, administrative, logistic) obtained from the each SNF shipment as from the Czech Republic as from other countries using the Russian type research reactors are evaluated and projected onto preparation of next shipment of high enriched nuclear fuel back to the Russian Federation. The results shown all shipments provided by the UJV Rez, a. s. in the frame of the GTRI Program have been performed successfully and safely. It is expected the experience and results will be applied to preparation and completing of the Chinese Miniature Neutron Source Reactors (MNSR) Spent Nuclear Fuel Repatriation in the near future. (authors)« less

Application of Radiation Chemistry to Some Selected Technological Issues Related to the Development of Nuclear Energy.

PubMed

Bobrowski, Krzysztof; Skotnicki, Konrad; Szreder, Tomasz

2016-10-01

The most important contributions of radiation chemistry to some selected technological issues related to water-cooled reactors, reprocessing of spent nuclear fuel and high-level radioactive wastes, and fuel evolution during final radioactive waste disposal are highlighted. Chemical reactions occurring at the operating temperatures and pressures of reactors and involving primary transients and stable products from water radiolysis are presented and discussed in terms of the kinetic parameters and radiation chemical yields. The knowledge of these parameters is essential since they serve as input data to the models of water radiolysis in the primary loop of light water reactors and super critical water reactors. Selected features of water radiolysis in heterogeneous systems, such as aqueous nanoparticle suspensions and slurries, ceramic oxides surfaces, nanoporous, and cement-based materials, are discussed. They are of particular concern in the primary cooling loops in nuclear reactors and long-term storage of nuclear waste in geological repositories. This also includes radiation-induced processes related to corrosion of cladding materials and copper-coated iron canisters, dissolution of spent nuclear fuel, and changes of bentonite clays properties. Radiation-induced processes affecting stability of solvents and solvent extraction ligands as well oxidation states of actinide metal ions during recycling of the spent nuclear fuel are also briefly summarized.

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

Resolving Past Liabilities for Future Reduction in Greenhouse Gases; Nuclear Energy and the Outstanding Federal Liability of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Donohue, Jay

This thesis will: (1) examine the current state of nuclear power in the U.S.; (2) provide a comparison of nuclear power to both existing alternative/renewable sources of energy as well as fossil fuels; (3) dissect Standard Contracts created pursuant to the National Waste Policy Act (NWPA), Congress' attempt to find a solution for Spent Nuclear Fuel (SNF), and the designation of Yucca Mountain as a repository; (4) the anticipated failure of Yucca Mountain; (5) explore WIPP as well as attempts to build a facility on Native American land in Utah; (6) examine reprocessing as a solution for SNF used by France and Japan; and, finally, (7) propose a solution to reduce GHG's by developing new nuclear energy plants with financial support from the U.S. government and a solution to build a storage facility for SNF through the sitting of a repository based on a "bottom-up" cooperative federalism approach.

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

Minor Actinides-Loaded FBR Core Concept Suitable for the Introductory Period in Japan

NASA Astrophysics Data System (ADS)

Fujimura, Koji; Sasahira, Akira; Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi

According to the Japan's Framework for Nuclear Energy Policy(1), a basic scenario for fast breeder reactors (FBRs) is that they will be introduced on a commercial basis starting around 2050 replacing light water reactors (LWRs). During the FBR introduction period, the Pu from LWR spent fuel is used for FBR startup. Howerver, the FBR core loaded with this Pu has a larger burnup reactivity due to its larger isotopic content of Pu-241 than a core loaded with Pu from an FBR multi-recycling core. The increased burnup reactivity may reduce the cycle length of an FBR. We investigated, an FBR transitional core concept to confront the issues of the FBR introductory period in Japan. Core specifications are based on the compact-type sodium-cooled mixed oxide (MOX)-fueled core designed from the Japanese FBR cycle feasibility studies, because lower Pu inventory should be better for the FBR introductory period in view of its flexibility for the required reprocessing amount of LWR spent fuel to start up FBRs. The reference specifications were selected as follows. Output of 1500MWe and average discharge fuel burnup of about 150GWd/t. Minor Actinides (MAs) recovered from LWR spent fuels which provide Pu to startup FBRs are loaded to the initial loading fuels and exchanged fuels during few cycles until equilibrium. We made the MA content of the initial loading fuel four kinds like 0%, 3%, 4%, 5%. The average of the initial loading fuel is assumed to be 3%, and that of the exchange fuel is set as 5%. This 5% maximum of the MA content is based on the irradiation results of the experimental fast reactor Joyo. We evaluated the core performances including burnup characteristics and the reactivity coefficient and confirmed that transitional core from initial loading until equilibrium cycle with loaded Pu from LWR spent fuel performs similary to an FBR multi-recycling core.

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.

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.

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.

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

Returning HEU Fuel from the Czech Republic to Russia

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

Michael Tyacke; Dr. Igor Bolshinsky

In December 1999, representatives from the United States, Russian Federation, and International Atomic Energy Agency began working on a program to return Russian supplied, highly enriched, uranium fuel stored at foreign research reactors to Russia. Now, under the Global Threat Reduction Initiative’s Russian Research Reactor Fuel Return Program, this effort has repatriated over 800 kg of highly enriched uranium to Russia from over 10 countries. In May 2004, the “Agreement Between the Government of the United States of America and the Government of the Russian Federation Concerning Cooperation for the Transfer of Russian Produced Research Reactor Nuclear Fuel to themore » Russian Federation” was signed. This agreement provides legal authority for the Russian Research Reactor Fuel Return Program and establishes parameters whereby eligible countries may return highly enriched uranium spent and fresh fuel assemblies and other fissile materials to Russia. On December 8, 2007, one of the largest shipments of highly 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. In February 2003, Russian Research Reactor Fuel Return Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their highly enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This article discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.« less

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

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

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.

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

Engineered Materials for Cesium and Strontium Storage Final Technical Report

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

Sean M. McDeavitt

2010-04-14

Closing the nuclear fuel cycle requires reprocessing spent fuel to recover the long-lived components that still have useful energy content while immobilizing the remnant waste fission products in stable forms. At the genesis of this project, next generation spent fuel reprocessing methods were being developed as part of the U.S. Department of Energy's Advanced Fuel Cycle Initiative. One of these processes was focused on solvent extraction schemes to isolate cesium (Cs) and strontium (Sr) from spent nuclear fuel. Isolating these isotopes for short-term decay storage eases the design requirements for long-term repository disposal; a significant amount of the radiation andmore » decay heat in fission product waste comes from Cs-137 and Sr-90. For the purposes of this project, the Fission Product Extraction (FPEX) process is being considered to be the baseline extraction method. The objective of this project was to evaluate the nature and behavior of candidate materials for cesium and strontium immobilization; this will include assessments with minor additions of yttrium, barium, and rubidium in these materials. More specifically, the proposed research achieved the following objectives (as stated in the original proposal): (1) Synthesize simulated storage ceramics for Cs and Sr using an existing labscale steam reformer at Purdue University. The simulated storage materials will include aluminosilicates, zirconates and other stable ceramics with the potential for high Cs and Sr loading. (2) Characterize the immobilization performance, phase structure, thermal properties and stability of the simulated storage ceramics. The ceramic products will be stable oxide powders and will be characterized to quantify their leach resistance, phase structure, and thermophysical properties. The research progressed in two stages. First, a steam reforming process was used to generate candidate Cs/Sr storage materials for characterization. This portion of the research was carried out at Purdue University and is detailed in Appendix A. Steam reforming proved to be too rigorous for efficient The second stage of this project was carried out at Texas A&M University and is Detailed in Appendix B. In this stage, a gentler ceramic synthesis process using Cs and Sr loaded kaolinite and bentonite clays was developed in collaboration with Dr. M. Kaminski at Argonne National Laboratory.« less

Experimental study on thermal hazard of tributyl phosphate-nitric acid mixtures using micro calorimeter technique.

PubMed

Sun, Qi; Jiang, Lin; Gong, Liang; Sun, Jin-Hua

2016-08-15

During PUREX spent nuclear fuel reprocessing, mixture of tributyl phosphate (TBP) and hydrocarbon solvent are employed as organic solvent to extract uranium in consideration of radiation contaminated safety and resource recycling, meanwhile nitric acid is utilized to dissolve the spent fuel into small pieces. However, once TBP contacts with nitric acid or nitrates above 130°C, a heavy "red oil" layer would occur accompanied by thermal runaway reactions, even caused several nuclear safety accident. Considering nitric acid volatility and weak exothermic detection, C80micro calorimeter technique was used in this study to investigate thermal decomposition of TBP mixed with nitric acid. Results show that the concentration of nitric acid greatly influences thermal hazard of the system by direct reactions. Even with a low heating rate, if the concentration of nitric acid increases due to evaporation of water or improper operations, thermal runaway in the closed system could start at a low temperature. Copyright © 2016 Elsevier B.V. All rights reserved.

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Stacey, W. M.

2009-09-01

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

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.

Cigeo, the French Geological Repository Project - 13022

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

Labalette, Thibaud; Harman, Alain; Dupuis, Marie-Claude

The Cigeo industrial-scale geological disposal centre is designed for the disposal of the most highly-radioactive French waste. It will be built in an argillite formation of the Callovo-Oxfordian dating back 160 million years. The Cigeo project is located near the Bure village in the Paris Basin. The argillite formation was studied since 1974, and from the Meuse/Haute-Marne underground research laboratory since end of 1999. Most of the waste to be disposed of in the Cigeo repository comes from nuclear power plants and from reprocessing of their spent fuel. (authors)

Pyrochemical and Dry Processing Methods Program. A selected bibliography

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

McDuffie, H.F.; Smith, D.H.; Owen, P.T.

1979-03-01

This selected bibliography with abstracts was compiled to provide information support to the Pyrochemical and Dry Processing Methods (PDPM) Program sponsored by DOE and administered by the Argonne National Laboratory. Objectives of the PDPM Program are to evaluate nonaqueous methods of reprocessing spent fuel as a route to the development of proliferation-resistant and diversion-resistant methods for widespread use in the nuclear industry. Emphasis was placed on the literature indexed in the ERDA--DOE Energy Data Base (EDB). The bibliography includes indexes to authors, subject descriptors, EDB subject categories, and titles.

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.

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.

Planning, Preparation, and Transport of the High-Enriched Uranium Spent Nuclear Fuel from the Czech Republic to the Russian Federation

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

M. J. Tyacke; I. Bolshinsky; Frantisek Svitak

The United States, Russian Federation, and the International Atomic Energy Agency have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program, which is part of the Global Threat Reduction Initiative. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. In February 2003, the RRRFR Program began discussions with the Nuclear Research Institute (NRI) in Rež, Czech Republic, about returning their HEU spent nuclear fuel to the Russian Federation for reprocessing. In March 2005, the U.S. Department ofmore » Energy signed a contract with NRI to perform all activities needed for transporting their HEU spent nuclear fuel to Russia. After 2 years of intense planning, preparations, and coordination at NRI and with three other countries, numerous organizations and agencies, and a Russian facility, this shipment is scheduled for completion before the end of 2007. This paper will provide a summary of activities completed for making this international shipment. This paper contains an introduction and background of the RRRFR Program and the NRI shipment project. It summarizes activities completed in preparation for the shipment, including facility preparations at NRI in Rež and FSUE “Mayak” in Ozyorsk, Russia; a new transportation cask system; regulatory approvals; transportation planning and preparation in the Czech Republic, Slovakia, Ukraine, and the Russian Federation though completion of the Unified Project and Special Ecological Programs. The paper also describes fuel loading and cask preparations at NRI and final preparations/approvals for transporting the shipment across the Czech Republic, Slovakia, Ukraine, and the Russian Federation to FSUE Mayak where the HEU spent nuclear fuel will be processed, the uranium will be downblended and made into low-enriched uranium fuel for commercial reactor use, and the high-level waste from the processing will be stabilized and stored for less than 20 years before being sent back to the Czech Republic for final disposition. Finally, the paper contains a section for the summary and conclusions.« less

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.

Carbon bed mercury emissions control for mixed waste treatment.

PubMed

Soelberg, Nick; Enneking, Joe

2010-11-01

Mercury has various uses in nuclear fuel reprocessing and other nuclear processes, and so it is often present in radioactive and mixed (radioactive and hazardous) wastes. Compliance with air emission regulations such as the Hazardous Waste Combustor (HWC) Maximum Achievable Control Technology (MACT) standards can require off-gas mercury removal efficiencies up to 99.999% for thermally treating some mixed waste streams. Test programs have demonstrated this level of off-gas mercury control using fixed beds of granular sulfur-impregnated activated carbon. Other results of these tests include (1) the depth of the mercury control mass transfer zone was less than 15-30 cm for the operating conditions of these tests; (2) MERSORB carbon can sorb mercury up to 19 wt % of the carbon mass; and (3) the spent carbon retained almost all (98.3-99.99%) of the mercury during Toxicity Characteristic Leachability Procedure (TCLP) tests, but when even a small fraction of the total mercury dissolves, the spent carbon can fail the TCLP test when the spent carbon contains high mercury concentrations.

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

Ibarra, Luis; Sanders, David; Yang, Haori

The main goal of this study is to evaluate the long-term seismic performance of freestanding and anchored Dry Storage Casks (DSCs) using experimental tests on a shaking table, as well as comprehensive numerical evaluations that include the cask-pad-soil system. The study focuses on the dynamic performance of vertical DSCs, which can be designed as free-standing structures resting on a reinforced concrete foundation pad, or casks anchored to a foundation pad. The spent nuclear fuel (SNF) at nuclear power plants (NPPs) is initially stored in fuel-storage pools to control the fuel temperature. After several years, the fuel assemblies are transferred tomore » DSCs at sites contiguous to the plant, known as Interim Spent Fuel Storage Installations (ISFSIs). The regulations for these storage systems (10 CFR 72) ensure adequate passive heat removal and radiation shielding during normal operations, off-normal events, and accident scenarios. The integrity of the DSCs is important, even if the overpack does not breach, because eventually the spent fuel-rods need to be shipped either to a reprocessing plant or a repository. DSCs have been considered as a temporary storage solution, and usually are licensed for 20 years, although they can be relicensed for operating periods of up to 60 years. In recent years, DSCs have been reevaluated as a potential mid-term solution, in which the operating period may be extended for up to 300 years. At the same time, recent seismic events have underlined the significant risks DSCs are exposed. The consideration of DCSs for storing spent fuel for hundreds of years has created new challenges. In the case of seismic hazard, longer-term operating periods not only lead to larger horizontal accelerations, but also increase the relative effect of vertical accelerations that usually are disregarded for smaller seismic events. These larger seismic demands could lead to casks sliding and tipping over, impacting the concrete pad or adjacent casks. The casks may also slide and collide with other casks or structural components. Also, the different DSC components may impact each other during these events. This study provides a comprehensive evaluation of DSCs subjected to these extreme demands, including the effect of vertical accelerations, and soilstructure interaction.« less

Zirconia-magnesia inert matrix fuel and waste form: Synthesis, characterization and chemical performance in an advanced fuel cycle

NASA Astrophysics Data System (ADS)

Holliday, Kiel Steven

There is a significant buildup in plutonium stockpiles throughout the world, because of spent nuclear fuel and the dismantling of weapons. The radiotoxicity of this material and proliferation risk has led to a desire for destroying excess plutonium. To do this effectively, it must be fissioned in a reactor as part of a uranium free fuel to eliminate the generation of more plutonium. This requires an inert matrix to volumetrically dilute the fissile plutonium. Zirconia-magnesia dual phase ceramic has been demonstrated to be a favorable material for this task. It is neutron transparent, zirconia is chemically robust, magnesia has good thermal conductivity and the ceramic has been calculated to conform to current economic and safety standards. This dissertation contributes to the knowledge of zirconia-magnesia as an inert matrix fuel to establish behavior of the material containing a fissile component. First, the zirconia-magnesia inert matrix is synthesized in a dual phase ceramic containing a fissile component and a burnable poison. The chemical constitution of the ceramic is then determined. Next, the material performance is assessed under conditions relevant to an advanced fuel cycle. Reactor conditions were assessed with high temperature, high pressure water. Various acid solutions were used in an effort to dissolve the material for reprocessing. The ceramic was also tested as a waste form under environmental conditions, should it go directly to a repository as a spent fuel. The applicability of zirconia-magnesia as an inert matrix fuel and waste form was tested and found to be a promising material for such applications.

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

Separation of mixtures of chemical elements in plasma

NASA Astrophysics Data System (ADS)

Dolgolenko, D. A.; Muromkin, Yu A.

2017-10-01

This paper reviews proposals on the plasma processing of radioactive waste (RW) and spent nuclear fuel (SNF). The chemical processing of SNF based on the extraction of its components from water solutions is rather expensive and produces new waste. The paper considers experimental research on plasma separation of mixtures of chemical elements and isotopes, whose results can help evaluate the plasma methods of RW and SNF reprocessing. The analysis identifies the difference between ionization levels of RW and SNF components at their transition to the plasma phase as a reason why all plasma methods are difficult to apply.

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

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.

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

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

Preparation, certification and validation of a stable solid spike of uranium and plutonium coated with a cellulose derivative for the measurement of uranium and plutonium content in dissolved nuclear fuel by isotope dilution mass spectrometry.

PubMed

Surugaya, Naoki; Hiyama, Toshiaki; Verbruggen, André; Wellum, Roger

2008-02-01

A stable solid spike for the measurement of uranium and plutonium content in nitric acid solutions of spent nuclear fuel by isotope dilution mass spectrometry has been prepared at the European Commission Institute for Reference Materials and Measurements in Belgium. The spike contains about 50 mg of uranium with a 19.838% (235)U enrichment and 2 mg of plutonium with a 97.766% (239)Pu abundance in each individual ampoule. The dried materials were covered with a thin film of cellulose acetate butyrate as a protective organic stabilizer to resist shocks encountered during transportation and to eliminate flaking-off during long-term storage. It was found that the cellulose acetate butyrate has good characteristics, maintaining a thin film for a long time, but readily dissolving on heating with nitric acid solution. The solid spike containing cellulose acetate butyrate was certified as a reference material with certified quantities: (235)U and (239)Pu amounts and uranium and plutonium amount ratios, and was validated by analyzing spent fuel dissolver solutions of the Tokai reprocessing plant in Japan. This paper describes the preparation, certification and validation of the solid spike coated with a cellulose derivative.

Regional and global contributions of anthropogenic iodine-129 in monthly deposition samples collected in North East Japan between 2006 and 2015.

PubMed

Hasegawa, Hidenao; Kakiuchi, Hideki; Akata, Naofumi; Ohtsuka, Yoshihito; Hisamatsu, Shun'ichi

2017-05-01

We measured the monthly atmospheric deposition flux of 129 I at Rokkasho, Aomori, Japan-the location of a commercial spent nuclear fuel reprocessing plant-from 2006 to 2015 to assess the impact of the plant on environmental 129 I levels. The plant is now under final safety assessment by a national authority after test operation using actual spent nuclear fuel. During cutting and chemical processing in test operations from April 2006 to October 2008, 129 I was discharged to the atmosphere and detected in our deposition samples. 129 I deposition fluxes largely followed the discharge pattern of 129 I from the plant to the atmosphere, and most of the deposited 129 I originated from the plant. In and after 2009, 129 I deposition fluxes decreased dramatically to reach the background level; the 129 I deposition fluxes at Rokkasho were almost the same as those at Hirosaki, where an additional sampling point was set up as a background site 85 km from the plant in 2011. The background 129 I deposition fluxes showed seasonal variation-high in winter and low in the other seasons-at both Rokkasho and Hirosaki. The results of a backward trajectory analysis of the air mass at Rokkasho suggested that reprocessing plants in Europe were the origins of the high 129 I flux in winter. The contribution of 129 I released from the Fukushima Dai-ichi Nuclear Power Plant accident to the 129 I deposition flux at Rokkasho in 2011 was small on the basis of the 129 I/ 131 I activity ratio. Copyright © 2017 Elsevier Ltd. All rights reserved.

Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

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

Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

1976-04-01

Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

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.

Water Chemistry Control System for Recovery of Damaged and Degraded Spent Fuel

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

Sindelar, R.; Fisher, D.; Thomas, J.

2011-02-18

The International Atomic Energy Agency (IAEA) and the government of Serbia have led the project cosponsored by the U.S, Russia, European Commission, and others to repackage and repatriate approximately 8000 spent fuel elements from the RA reactor fuel storage basins at the VIN?A Institute of Nuclear Sciences to Russia for reprocessing. The repackaging and transportation activities were implemented by a Russian consortium which includes the Sosny Company, Tekhsnabeksport (TENEX) and Mayak Production Association. High activity of the water of the fuel storage basin posed serious risk and challenges to the fuel removal from storage containers and repackaging for transportation. Themore » risk centered on personnel exposure, even above the basin water, due to the high water activity levels caused by Cs-137 leached from fuel elements with failed cladding. A team of engineers from the U.S. DOE-NNSA's Global Threat Reduction Initiative, the Vinca Institute, and the IAEA performed the design, development, and deployment of a compact underwater water chemistry control system (WCCS) to remove the Cs-137 from the basin water and enable personnel safety above the basin water for repackaging operations. Key elements of the WCCS system included filters, multiple columns containing an inorganic sorbent, submersible pumps and flow meters. All system components were designed to be remotely serviceable and replaceable. The system was assembled and successfully deployed at the Vinca basin to support the fuel removal and repackaging activities. Following the successful operations, the Cs-137 is now safely contained and consolidated on the zeolite sorbent used in the columns of the WCCS, and the fuel has been removed from the basins. This paper reviews the functional requirements, design, and deployment of the WCCS.« less

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.

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.

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.

Empirical model for calculating vapor-liquid equilibrium and associated phase enthalpy for the CO$sub 2$--O$sub 2$--Kr--Xe system for application to the KALC process

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

Glass, R. W.; Gilliam, T. M.; Fowler, V. L.

An empirical model is presented for vapor-liquid equilibria and enthalpy for the CO$sub 2$-O$sub 2$ system. In the model, krypton and xenon in very low concentrations are combined with the CO$sub 2$-O$sub 2$ system, thereby representing the total system of primary interest in the High-Temperature Gas- Cooled Reactor program for removing krypton from off-gas generated during the reprocessing of spent fuel. Selected properties of the individual and combined components being considered are presented in the form of tables and empirical equations. (auth)

Verification measurements of the IRMM-1027 and the IAEA large-sized dried (LSD) spikes.

PubMed

Jakopič, R; Aregbe, Y; Richter, S; Zuleger, E; Mialle, S; Balsley, S D; Repinc, U; Hiess, J

2017-01-01

In the frame of the accountancy measurements of the fissile materials, reliable determinations of the plutonium and uranium content in spent nuclear fuel are required to comply with international safeguards agreements. Large-sized dried (LSD) spikes of enriched 235 U and 239 Pu for isotope dilution mass spectrometry (IDMS) analysis are routinely applied in reprocessing plants for this purpose. A correct characterisation of these elements is a pre-requirement for achieving high accuracy in IDMS analyses. This paper will present the results of external verification measurements of such LSD spikes performed by the European Commission and the International Atomic Energy Agency.

Corrosion-Resistant Ti- xNb- xZr Alloys for Nitric Acid Applications in Spent Nuclear Fuel Reprocessing Plants

NASA Astrophysics Data System (ADS)

Manivasagam, Geetha; Anbarasan, V.; Kamachi Mudali, U.; Raj, Baldev

2011-09-01

This article reports the development, microstructure, and corrosion behavior of two new alloys such as Ti-4Nb-4Zr and Ti-2Nb-2Zr in boiling nitric acid environment. The corrosion test was carried out in the liquid, vapor, and condensate phases of 11.5 M nitric acid, and the potentiodynamic anodic polarization studies were performed at room temperature for both alloys. The samples subjected to three-phase corrosion testing were characterized using scanning electron microscopy (SEM) and energy-dispersive X-ray microanalysis (EDAX). As Ti-2Nb-2Zr alloy exhibited inferior corrosion behavior in comparison to Ti-4Nb-4Zr in all three phases, weldability and heat treatment studies were carried out only on Ti-4Nb-4Zr alloy. The weldability of the new alloy was evaluated using tungsten inert gas (TIG) welding processes, and the welded specimen was thereafter tested for its corrosion behavior in all three phases. The results of the present investigation revealed that the newly developed near alpha Ti-4Nb-4Zr alloy possessed superior corrosion resistance in all three phases and excellent weldability compared to conventional alloys used for nitric acid application in spent nuclear reprocessing plants. Further, the corrosion resistance of the beta heat-treated Ti-4Nb-4Zr alloy was superior when compared to the sample heat treated in the alpha + beta phase.

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

Hecker, Siegfried S.

Actions of the Government of the Democratic People's Republic of Korea have precipitated two nuclear crises in the past 10 years. The 1994 crisis was resolved through the 'Agreed Framework.' North Korea agreed to 'freeze' and eventually dismantle its nuclear program (with U.S. help to store spent fuel safely and under IAEA inspection). In return, the United States agreed (with the KEDO international consortium) to build two light-water reactors and supply North Korea with heavy-fuel oil until the reactors come on line. In addition, both sides agreed to move towards full normalization of relations, work for peace and security onmore » a nuclear-free Korean Peninsula, and work on strengthening the international nonproliferation regime. The second nuclear crisis erupted when North Korean Government officials allegedly admitted to having a clandestine uranium enrichment program when confronted with this accusation by U.S. officials in October 2002. The United States (through KEDO) suspended heavy-fuel oil shipments and North Korea responded by expelling the IAEA inspectors, withdrawing from the Nuclear Nonproliferation Treaty, and restarting its nuclear program in January 2003. The North Korean Government has invited Professor John Lewis of Stanford University, a China and North Korea scholar, for Track I1 discussions of nuclear and other key issues since 1987. In August 2003, Professor Lewis visited North Korea just before the first six-party talks, which were designed by the United States to solve the current nuclear crisis. Professor Lewis was invited back for the January 2004 visit. He asked Jack Pritchard, former U.S. special envoy for DRPK negotiations, and me to accompany him. Two Asian affairs staff specialists from the U.S. Senate Foreign Relations Committee also joined us. I will report on the visit to the Yongbyon Nuclear Scientific Research Center on January 8,2004. We toured the 5 MWe reactor, the 50 MWe reactor construction site, the spent fuel pool storage building, and the radiochemical laboratory. We concluded that North Korea has restarted its 5 MWe reactor (which produces roughly 6 kg of plutonium annually), it removed the 8000 spent fuel rods that were previously stored under IAEA safeguards from the spent fuel pool, and that it most likely extracted the 25 to 30 kg of plutonium contained in these fuel rods. Although North Korean officials showed us what they claimed was their plutonium metal product from this reprocessing campaign, we were not able to conclude definitively that it was in fact plutonium metal and that it came from the most recent reprocessing campaign. Nevertheless, our North Korean hosts demonstrated that they had the capability, the facility and requisite capacity, and the technical expertise to produce plutonium metal. We were not shown any facilities or had the opportunity to talk to technical or military experts who were able to address the issue of whether or not North Korea had a 'deterrent' as claimed - that is, we were not able to conclude that North Korea can build a nuclear device and that it can integrate nuclear devices into suitable delivery systems. On the matter of uranium enrichment programs, Vice Minister Kim Gye Gwan categorically denied that North Korea has a uranium enrichment program - he said, 'we have no program, no equipment, and no technical expertise for uranium enrichment.' Upon return to the United States, I shared my observations and analysis with U.S. Government officials in Washington, DC, including congressional testimony to the Senate Foreign Relations Committee and briefings to two House of Representative Committees.« less

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

Study of Compton suppression for use in spent nuclear fuel assay

NASA Astrophysics Data System (ADS)

Bender, Sarah

The focus of this study has been to assess Compton suppressed gamma-ray detection systems for the multivariate analysis of spent nuclear fuel. This objective has been achieved using direct measurement of samples of irradiated fuel elements in two geometrical configurations with Compton suppression systems. In order to address the objective to quantify the number of additionally resolvable photopeaks, direct Compton suppressed spectroscopic measurements of spent nuclear fuel in two configurations were performed: as intact fuel elements and as dissolved feed solutions. These measurements directly assessed and quantified the differences in measured gamma-ray spectrum from the application of Compton suppression. Several irradiated fuel elements of varying cooling time from the Penn State Breazeale Reactor spent fuel inventory were measured using three Compton suppression systems that utilized different primary detectors: HPGe, LaBr3, and NaI(Tl). The application of Compton suppression using a LaBr3 primary detector to the measurement of the current core fuel element, which presented the highest count rate, allowed four additional spectral features to be resolved. In comparison, the HPGe-CSS was able to resolve eight additional photopeaks as compared to the standalone HPGe measurement. Measurements with the NaI(Tl) primary detector were unable to resolve any additional peaks, due to its relatively low resolution. Samples of Approved Test Material (ATM) commercial fuel elements were obtained from Pacific Northwest National Laboratory. The samples had been processed using the beginning stages of the PUREX method and represented the unseparated feed solution from a reprocessing facility. Compton suppressed measurements of the ATM fuel samples were recorded inside the guard detector annulus, to simulate the siphoning of small quantities from the main process stream for long dwell measurement periods. Photopeak losses were observed in the measurements of the dissolved ATM fuel samples because the spectra was recorded from the source in very close proximity to the detector and surrounded by the guard annulus, so the detection probability is very high. Though this configuration is optimal for a Compton suppression system for the measurement of low count rate samples, measurement of high count rate samples in the enclosed arrangement leads to sum peaks in both the suppressed and unsuppressed spectra and losses to photopeak counts in the suppressed spectra. No additional photopeaks were detected using Compton suppression with this geometry. A detector model was constructed that can accurately simulate a Compton suppressed spectral measurement of radiation from spent nuclear fuel using HPGe or LaBr3 detectors. This is the first detector model capable of such an accomplishment. The model uses the Geant4 toolkit coupled with the RadSrc application and it accepts spent fuel composition data in list form. The model has been validated using dissolved ATM fuel samples in the standard, enclosed geometry of the PSU HPGe-CSS. The model showed generally good agreement with both the unsuppressed and suppressed measured fuel sample spectra, however the simulation is more appropriate for the generation of gamma-ray spectra in the beam source configuration. Photopeak losses due to cascade decay emissions in the Compton suppressed spectra were not appropriately managed by the simulation. Compton suppression would be a beneficial addition to NDA process monitoring systems if oriented such that the gamma-ray photons are collimated to impinge the primary detector face as a beam. The analysis has shown that peak losses through accidental coincidences are minimal and the reduction in the Compton continuum allows additional peaks to be resolved. (Abstract shortened by UMI.).

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

Dunn, Darrell; Poinssot, Christophe; Begg, Bruce

Management of nuclear waste remains an important international topic that includes reprocessing of commercial nuclear fuel, waste-form design and development, storage and disposal packaging, the process of repository site selection, system design, and performance assessment. Requirements to manage and dispose of materials from the production of nuclear weapons, and the renewed interest in nuclear power, in particular through the Generation IV Forum and the Advanced Fuel Cycle Initiative, can be expected to increase the need for scientific advances in waste management. A broad range of scientific and engineering disciplines is necessary to provide safe and effective solutions and address complexmore » issues. This volume offers an interdisciplinary perspective on materials-related issues associated with nuclear waste management programs. Invited and contributed papers cover a wide range of topics including studies on: spent fuel; performance assessment and models; waste forms for low- and intermediate-level waste; ceramic and glass waste forms for plutonium and high-level waste; radionuclides; containers and engineered barriers; disposal environments and site characteristics; and partitioning and transmutation.« less

Molten salts and nuclear energy production

NASA Astrophysics Data System (ADS)

Le Brun, Christian

2007-01-01

Molten salts (fluorides or chlorides) were considered near the beginning of research into nuclear energy production. This was initially due to their advantageous physical and chemical properties: good heat transfer capacity, radiation insensitivity, high boiling point, wide range solubility for actinides. In addition it was realised that molten salts could be used in numerous situations: high temperature heat transfer, core coolants with solid fuels, liquid fuel in a molten salt reactor, solvents for spent nuclear solid fuel in the case of pyro-reprocessing and coolant and tritium production in the case of fusion. Molten salt reactors, one of the six innovative concepts chosen by the Generation IV international forum, are particularly interesting for use as either waste incinerators or thorium cycle systems. As the neutron balance in the thorium cycle is very tight, the possibility to perform online extraction of some fission product poisons from the salt is very attractive. In this article the most important questions that must be addressed to demonstrate the feasibility of molten salt reactor will be reviewed.

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.

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

Reducing Actinide Production Using Inert Matrix Fuels

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

Deinert, Mark

2017-08-23

The environmental and geopolitical problems that surround nuclear power stem largely from the longlived transuranic isotopes of Am, Cm, Np and Pu that are contained in spent nuclear fuel. New methods for transmuting these elements into more benign forms are needed. Current research efforts focus largely on the development of fast burner reactors, because it has been shown that they could dramatically reduce the accumulation of transuranics. However, despite five decades of effort, fast reactors have yet to achieve industrial viability. A critical limitation to this, and other such strategies, is that they require a type of spent fuel reprocessingmore » that can efficiently separate all of the transuranics from the fission products with which they are mixed. Unfortunately, the technology for doing this on an industrial scale is still in development. In this project, we explore a strategy for transmutation that can be deployed using existing, current generation reactors and reprocessing systems. We show that use of an inert matrix fuel to recycle transuranics in a conventional pressurized water reactor could reduce overall production of these materials by an amount that is similar to what is achievable using proposed fast reactor cycles. Furthermore, we show that these transuranic reductions can be achieved even if the fission products are carried into the inert matrix fuel along with the transuranics, bypassing the critical separations hurdle described above. The implications of these findings are significant, because they imply that inert matrix fuel could be made directly from the material streams produced by the commercially available PUREX process. Zirconium dioxide would be an ideal choice of inert matrix in this context because it is known to form a stable solid solution with both fission products and transuranics.« 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.

Evaluation of N,N-dialkylamides as promising process extractants

NASA Astrophysics Data System (ADS)

Pathak, P. N.; Prabhu, D. R.; Kanekar, A. S.; Manchanda, V. K.

2010-03-01

Studies carried out at BARC, India on the development of new extractants for reprocessing of spent fuel suggested that while straight chain N,N-dihexyloctanamide (DHOA) is promising alternative to TBP for the reprocessing of irradiated uranium based fuels, branched chain N,N-di(2-ethylhexyl)isobutyramide (D2EHIBA) is suitable for the selective recovery of 233U from irradiated Th. In advanced fuel cycle scenarios, the coprocessing of U/Pu stream appears attractive particularly with respect to development of proliferation resistant technologies. DHOA extracted Pu(IV) more efficiently than TBP, both at trace-level concentration as well as under uranium/plutonium loading conditions. Uranium extraction behavior of DHOA was however, similar to that of TBP during the extraction cycle. Stripping behavior of U and Pu (without any reductant) was better for DHOA than that of TBP. It was observed during batch studies that whereas 99% Pu is stripped in four stages in case of DHOA, only 89% Pu is stripped in case of TBP under identical experimental conditions. DHOA offered better fission product decontamination than that of TBP. GANEX (Group ActiNide EXtraction) and ARTIST (Amide-based Radio-resources Treatment with Interim Storage of Transuranics) processes proposed for actinide partitioning use branched chain amides for the selective extraction of uranium from spent fuel feed solutions. The branched-alkyl monoamide (BAMA) proposed to be used in ARTIST process is N,N-di-(2-ethylhexyl)butyramide (D2EHBA). In this context, the extraction behavior of U(VI) and Pu(IV) were compared using D2EHIBA, TBP, and D2EHBA under similar concentration of nitric acid (0.5 — 6M) and of uranium (0-50g/L). These studies suggested that D2EHIBA is a promising extractant for selective extraction of uranium over plutonium in process streams. Similarly, D2EHIBA offered distinctly better decontamination of 233U over Th and fission products under THOREX feed conditions. The possibility of simultaneous stripping and precipitation of thorium (as oxalate) from loaded organic phase was explored using 0.05M oxalic acid. Ammonium diuranate (ADU) precipitation was performed on the oxalate supernatant for the recovery of uranium. Quantitative recovery (>99.9%) of Th as well as of U was achieved. Radiolytic studies suggested that irradiated DHOA and D2EHIBA behaved better with respect to fission product decontamination as compared to that of TBP.

Insight into the Extraction Mechanism of Americium(III) over Europium(III) with Pyridylpyrazole: A Relativistic Quantum Chemistry Study.

PubMed

Kong, Xiang-He; Wu, Qun-Yan; Wang, Cong-Zhi; Lan, Jian-Hui; Chai, Zhi-Fang; Nie, Chang-Ming; Shi, Wei-Qun

2018-05-10

Separation of trivalent actinides (An(III)) and lanthanides (Ln(III)) is one of the most important steps in spent nuclear fuel reprocessing. However, it is very difficult and challenging to separate them due to their similar chemical properties. Recently the pyridylpyrazole ligand (PypzH) has been identified to show good separation ability toward Am(III) over Eu(III). In this work, to explore the Am(III)/Eu(III) separation mechanism of PypzH at the molecular level, the geometrical structures, bonding nature, and thermodynamic behaviors of the Am(III) and Eu(III) complexes with PypzH ligands modified by alkyl chains (Cn-PypzH, n = 2, 4, 8) have been systematically investigated using scalar relativistic density functional theory (DFT). According to the NBO (natural bonding orbital) and QTAIM (quantum theory of atoms in molecules) analyses, the M-N bonds exhibit a certain degree of covalent character, and more covalency appears in Am-N bonds compared to Eu-N bonds. Thermodynamic analyses suggest that the 1:1 extraction reaction, [M(NO 3 )(H 2 O) 6 ] 2+ + PypzH + 2NO 3 - → M(PypzH)(NO 3 ) 3 (H 2 O) + 5H 2 O, is the most suitable for Am(III)/Eu(III) separation. Furthermore, the extraction ability and the Am(III)/Eu(III) selectivity of the ligand PypzH is indeed enhanced by adding alkyl-substituted chains in agreement with experimental observations. Besides this, the nitrogen atom of pyrazole ring plays a more significant role in the extraction reactions related to Am(III)/Eu(III) separation compared to that of pyridine ring. This work could identify the mechanism of the Am(III)/Eu(III) selectivity of the ligand PypzH and provide valuable theoretical information for achieving an efficient Am(III)/Eu(III) separation process for spent nuclear fuel reprocessing.

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

Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

Worldwide activities related to nuclear fuel cycle and radioactive waste management programs are summarized. Several trends have developed in waste management strategy: All countries having to dispose of reprocessing wastes plan on conversion of the high-level waste (HLW) stream to a borosilicate glass and eventual emplacement of the glass logs, suitably packaged, in a deep geologic repository. Countries that must deal with plutonium-contaminated waste emphasize pluonium recovery, volume reduction and fixation in cement or bitumen in their treatment plans and expect to use deep geologic repositories for final disposal. Commercially available, classical engineering processing are being used worldwide to treatmore » and immobilize low- and intermediate-level wastes (LLW, ILW); disposal to surface structures, shallow-land burial and deep-underground repositories, such as played-out mines, is being done widely with no obvious technical problems. Many countries have established extensive programs to prepare for construction and operation of geologic repositories. Geologic media being studied fall into three main classes: argillites (clay or shale); crystalline rock (granite, basalt, gneiss or gabbro); and evaporates (salt formations). Most nations plan to allow 30 years or longer between discharge of fuel from the reactor and emplacement of HLW or spent fuel is a repository to permit thermal and radioactive decay. Most repository designs are based on the mined-gallery concept, placing waste or spent fuel packages into shallow holes in the floor of the gallery. Many countries have established extensive and costly programs of site evaluation, repository development and safety assessment. Two other waste management problems are the subject of major R and D programs in several countries: stabilization of uranium mill tailing piles; and immobilization or disposal of contaminated nuclear facilities, namely reactors, fuel cycle plants and R and D laboratories.« less

The ``Nuclear Renaissance'' and the Spread of Nuclear Weapons

NASA Astrophysics Data System (ADS)

Lyman, Edwin S.

2007-05-01

As interest grows around the world in nuclear power as an energy source that could help control greenhouse gas emissions, some have proclaimed the arrival of a ``nuclear renaissance.'' But can the increased risks of more nuclear power be managed? The political crisis surrounding Iran's pursuit of uranium enrichment has exposed weaknesses in the nuclear nonproliferation regime. Also, al Qaeda's declared interest in weapons of mass destruction raises the concern that terrorists could acquire nuclear weapons by stealing materials from poorly secured facilities. Growth of nuclear energy would require the construction of many additional uranium enrichment plants. And the generation of more spent nuclear fuel without a credible waste disposal strategy would increase political support for reprocessing, which separates large quantities of weapon-usable plutonium from spent fuel. There is little evidence that the various institutional arrangements and technical schemes proposed to mitigate the security risks of a major nuclear expansion would be effective. This talk will focus on the measures necessary to allow large-scale global growth of nuclear power without resulting in an unacceptably high risk of nuclear proliferation and nuclear terrorism, and will discuss the feasibility of such measures. To cite this abstract, use the following reference: http://meetings.aps.org/link/BAPS.2007.OSS07.E1.2

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.

Recent advances in computational actinoid chemistry.

PubMed

Wang, Dongqi; van Gunsteren, Wilfred F; Chai, Zhifang

2012-09-07

We briefly review advances in computational actinoid (An) chemistry during the past ten years in regard to two issues: the geometrical and electronic structures, and reactions. The former addresses the An-O, An-C, and M-An (M is a metal atom including An) bonds in the actinoid molecular systems, including actinoid oxo and oxide species, actinoid-carbenoid, dinuclear and diatomic systems, and the latter the hydration and ligand exchange, the disproportionation, the oxidation, the reduction of uranyl, hydroamination, and the photolysis of uranium azide. Concerning their relevance to the electronic structures and reactions of actinoids and their importance in the development of an advanced nuclear fuel cycle, we also mentioned the work on actinoid carbides and nitrides, which have been proposed to be candidates of the next generation of nuclear fuel, and the oxidation of PuO(x), which is important to understand the speciation of actinoids in the environment, followed by a brief discussion on the urgent need for a heavier involvement of computational actinoid chemistry in developing advanced reprocessing protocols of spent nuclear fuel. The paper is concluded with an outlook.

Eddy Current for Sizing Cracks in Canisters for Dry Storage of Used Nuclear Fuel

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

Meyer, Ryan M.; Jones, Anthony M.; Pardini, Allan F.

2014-01-01

The storage of used nuclear fuel (UNF) in dry canister storage systems (DCSSs) at Independent Spent Fuel Storage Installations (ISFSI) sites is a temporary measure to accommodate UNF inventory until it can be reprocessed or transferred to a repository for permanent disposal. Policy uncertainty surrounding the long-term management of UNF indicates that DCSSs will need to store UNF for much longer periods than originally envisioned. Meanwhile, the structural and leak-tight integrity of DCSSs must not be compromised. The eddy current technique is presented as a potential tool for inspecting the outer surfaces of DCSS canisters for degradation, particularly atmospheric stressmore » corrosion cracking (SCC). Results are presented that demonstrate that eddy current can detect flaws that cannot be detected reliably using standard visual techniques. In addition, simulations are performed to explore the best parameters of a pancake coil probe for sizing of SCC flaws in DCSS canisters and to identify features in frequency sweep curves that may potentially be useful for facilitating accurate depth sizing of atmospheric SCC flaws from eddy current measurements.« less

Solid wastes from nuclear power production.

PubMed Central

Soule, H F

1978-01-01

Radioactivity in nuclear power effluents is negligible compared to that in retained wastes to be disposed of as solids. Two basic waste categories are those for which shallow disposal is accepted and those for which more extreme isolation is desired. The latter includes "high level" wastes and others contaminated with radionuclides with the unusual combined properties of long radioactive half-life and high specific radiotoxicity. The favored method for extreme isolation is emplacement in a deep stable geologic formation. Necessary technologies for waste treatment and disposal are considered available. The present program to implement these technologies is discussed, including the waste management significance of current policy on spent nuclear fuel reprocessing. Recent difficulties with shallow disposal of waste are summarized. PMID:738244

Closed Fuel Cycle Waste Treatment Strategy

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

Vienna, J. D.; Collins, E. D.; Crum, J. V.

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significantmore » additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.« less

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.)

Endoscope reprocessing methods: a prospective study on the impact of human factors and automation.

PubMed

Ofstead, Cori L; Wetzler, Harry P; Snyder, Alycea K; Horton, Rebecca A

2010-01-01

The main cause of endoscopy-associated infections is failure to adhere to reprocessing guidelines. More information about factors impacting compliance is needed to support the development of effective interventions. The purpose of this multisite, observational study was to evaluate reprocessing practices, employee perceptions, and occupational health issues. Data were collected utilizing interviews, surveys, and direct observation. Written reprocessing policies and procedures were in place at all five sites, and employees affirmed the importance of most recommended steps. Nevertheless, observers documented guideline adherence, with only 1.4% of endoscopes reprocessed using manual cleaning methods with automated high-level disinfection versus 75.4% of those reprocessed using an automated endoscope cleaner and reprocessor. The majority reported health problems (i.e., pain, decreased flexibility, numbness, or tingling). Physical discomfort was associated with time spent reprocessing (p = .041). Discomfort diminished after installation of automated endoscope cleaners and reprocessors (p = .001). Enhanced training and accountability, combined with increased automation, may ensure guideline adherence and patient safety while improving employee satisfaction and health.

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.

High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

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

Mcwilliams, A. J.

2015-09-08

This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniquesmore » through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.« less

Superconducting RF Linacs Driving Subcritical Reactors for Profitable Disposition of Surplus Weapons-grade Plutonium

NASA Astrophysics Data System (ADS)

Cummings, Mary Anne; Johnson, Rolland

Acceptable capital and operating costs of high-power proton accelerators suitable for profitable commercial electric-power and process-heat applications have been demonstrated. However, studies have pointed out that even a few hundred trips of an accelerator lasting a few seconds would lead to unacceptable thermal stresses as each trip causes fission to be turned off in solid fuel structures found in conventional reactors. The newest designs based on the GEM*STAR concept take such trips in stride by using molten-salt fuel, where fuel pin fatigue is not an issue. Other aspects of the GEM*STAR concept which address all historical reactor failures include an internal spallation neutron target and high temperature molten salt fuel with continuous purging of volatile radioactive fission products such that the reactor contains less than a critical mass and almost a million times fewer volatile radioactive fission products than conventional reactors. GEM*STAR is a reactor that without redesign will burn spent nuclear fuel, natural uranium, thorium, or surplus weapons material. It will operate without the need for a critical core, fuel enrichment, or reprocessing making it an excellent candidate for export. As a first application, the design for a pilot plant is described for the profitable disposition of surplus weapons-grade plutonium by using process heat to produce green diesel fuel for the Department of Defense (DOD) from natural gas and renewable carbon.

A Computer Model of the Evaporator for the Development of an Automatic Control System

NASA Astrophysics Data System (ADS)

Kozin, K. A.; Efremov, E. V.; Kabrysheva, O. P.; Grachev, M. I.

2016-08-01

For the implementation of a closed nuclear fuel cycle it is necessary to carry out a series of experimental studies to justify the choice of technology. In addition, the operation of the radiochemical plant is impossible without high-quality automatic control systems. In the technologies of spent nuclear fuel reprocessing, the method of continuous evaporation is often used for a solution conditioning. Therefore, the effective continuous technological process will depend on the operation of the evaporation equipment. Its essential difference from similar devices is a small size. In this paper the method of mathematic simulation is applied for the investigation of one-effect evaporator with an external heating chamber. Detailed modelling is quite difficult because the phase equilibrium dynamics of the evaporation process is not described. Moreover, there is a relationship with the other process units. The results proved that the study subject is a MIMO plant, nonlinear over separate control channels and not selfbalancing. Adequacy was tested using the experimental data obtained at the laboratory evaporation unit.

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

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

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

2012-06-06

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

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.

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

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.

Radioactivity levels in major French rivers: summary of monitoring chronicles acquired over the past thirty years and current status.

PubMed

Eyrolle, Frédérique; Claval, David; Gontier, Gilles; Antonelli, Christelle

2008-07-01

Since the beginning of the 1990 s, liquid releases of gamma-emitting radionuclides from French nuclear facilities have generally fallen by almost 85%. Almost 65% of gamma-emitting liquid effluents released into freshwater rivers concerned the River Rhône (Southeast France), with around 85% of this originating from the Marcoule spent fuel reprocessing plant. Upstream of French nuclear plants, artificial radionuclides still detected by gamma spectrometry in 2006, include (137)Cs, (131)I as well as (60)Co, (58)Co and (54)Mn in the case of the Rhine (Switzerland nuclear industries). In the wake of the fallout from the Chernobyl accident, (103)Ru, (106)Rh-Ru, (110 m)Ag, (141)Ce and (129)Te were detected in rivers in the east of France. Some of these radionuclides were found in aquatic plants until 1989. In eastern France, (137)Cs activity in river sediments and mosses is still today two to three times greater than that observed in similar environments in western France. No (134)Cs has been detected upstream of nuclear plants in French rivers since 2001. Downstream of nuclear plants, the gamma emitters still detected regularly in rivers in 2006 are (137)Cs, (134)Cs, (60)Co, (58)Co, (110 m)Ag, (54)Mn, (131)I, together with (241)Am downstream of the Marcoule spent fuel reprocessing plant. Alpha and beta emitters such as plutonium isotopes and (90)Sr first entered freshwaters at the early 1950s due to the leaching of soils contaminated by atmospheric fallout from nuclear testing. These elements were also introduced, in the case of the Rhône River, via effluent from the Marcoule reprocessing plant. Until the mid 1990 s, plutonium isotope levels observed in the lower reaches of the Rhône were 10 to 1000 times higher than those observed in other French freshwaters. Data gathered over a period of almost thirty years of radioecological studies reveal that the only radionuclides detected in fish muscles are (137)Cs, (90)Sr, plutonium isotopes and (241)Am. At the scale of the French territory, there is no significant difference since the mid 1990 s between (137)Cs activity observed downstream of nuclear facilities and that observed upstream, whether in sediments, mosses and fish. Finally, this study highlights that the natural radioactivity of surface freshwaters are around 25 times greater than artificial radioactivity from gamma emitters. However, non gamma emitters released by nuclear industries, such as (3)H, may lead to artificial activity levels 2 to 20 times higher than natural levels.

Reforming of fuel inside fuel cell generator

DOEpatents

Grimble, Ralph E.

1988-01-01

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

Reforming of fuel inside fuel cell generator

DOEpatents

Grimble, R.E.

1988-03-08

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

Next-generation purex flowsheets with acetohydroxamic acid as complexant for FBR and thermal-fuel reprocessing

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

Kumar, Shekhar; Koganti, S.B.

2008-07-01

Acetohydroxamic acid (AHA) is a novel complexant for recycle of nuclear-fuel materials. It can be used in ordinary centrifugal extractors, eliminating the need for electro-redox equipment or complex maintenance requirements in a remotely maintained hot cell. In this work, the effect of AHA on Pu(IV) distribution ratios in 30% TBP system was quantified, modeled, and integrated in SIMPSEX code. Two sets of batch experiments involving macro Pu concentrations (conducted at IGCAR) and one high-Pu flowsheet (literature) were simulated for AHA based U-Pu separation. Based on the simulation and validation results, AHA based next-generation reprocessing flowsheets are proposed for co-processing basedmore » FBR and thermal-fuel reprocessing as well as evaporator-less macro-level Pu concentration process required for MOX fuel fabrication. Utilization of AHA results in significant simplification in plant design and simpler technology implementations with significant cost savings. (authors)« less

Uranium, its impact on the national and global energy mix; and its history, distribution, production, nuclear fuel-cycle, future, and relation to the environment

USGS Publications Warehouse

Finch, Warren Irvin

1997-01-01

The many aspects of uranium, a heavy radioactive metal used to generate electricity throughout the world, are briefly described in relatively simple terms intended for the lay reader. An adequate glossary of unfamiliar terms is given. Uranium is a new source of electrical energy developed since 1950, and how we harness energy from it is explained. It competes with the organic coal, oil, and gas fuels as shown graphically. Uranium resources and production for the world are tabulated and discussed by country and for various energy regions in the United States. Locations of major uranium deposits and power reactors in the United States are mapped. The nuclear fuel-cycle of uranium for a typical light-water reactor is illustrated at the front end-beginning with its natural geologic occurrence in rocks through discovery, mining, and milling; separation of the scarce isotope U-235, its enrichment, and manufacture into fuel rods for power reactors to generate electricity-and at the back end-the reprocessing and handling of the spent fuel. Environmental concerns with the entire fuel cycle are addressed. The future of the use of uranium in new, simplified, 'passively safe' reactors for the utility industry is examined. The present resource assessment of uranium in the United States is out of date, and a new assessment could aid the domestic uranium industry.

Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

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

Hadgu, Teklu; Hardin, Ernest; Matteo, Edward N.

2015-12-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are keptmore » open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

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

Hardin, Ernest; Matteo, Edward N.; Hadgu, Teklu

2016-01-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement formore » extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

Delayed gamma-ray spectroscopy with lanthanum bromide detector for non-destructive assay of nuclear material

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

Favalli, Andrea; Iliev, Metodi; Ianakiev, Kiril

High-energy delayed γ-ray spectroscopy is a potential technique for directly assaying spent fuel assemblies and achieving the safeguards goal of quantifying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Requirements for the γ-ray detection system, up to ~6 MeV, can be summarized as follows: high efficiency at high γ-ray energies, high energy resolution, good linearity between γ-ray energy and output signal amplitude, ability to operate at very high count rates, and ease of use in industrial environments such as nuclear facilities. High Purity Germanium Detectors (HPGe) are the state of the artmore » and provide excellent energy resolution but are limited in their count rate capability. Lanthanum Bromide (LaBr 3) scintillation detectors offer significantly higher count rate capabilities at lower energy resolution. Thus, LaBr 3 detectors may be an effective alternative for nuclear spent-fuel applications, where count-rate capability is a requirement. This paper documents the measured performance of a 2” (length) × 2” (diameter) of LaBr3 scintillation detector system, coupled to a negatively biased PMT and a tapered active high voltage divider, with count-rates up to ~3 Mcps. An experimental methodology was developed that uses the average current from the PMT’s anode and a dual source method to characterize the detector system at specific very high count rate values. Delayed γ-ray spectra were acquired with the LaBr 3 detector system at the Idaho Accelerator Center, Idaho State University, where samples of ~3g of 235U were irradiated with moderated neutrons from a photo-neutron source. Results of the spectroscopy characterization and analysis of the delayed γ-ray spectra acquired indicate the possible use of LaBr3 scintillation detectors when high count rate capability may outweigh the lower energy resolution.« less

Delayed gamma-ray spectroscopy with lanthanum bromide detector for non-destructive assay of nuclear material

DOE PAGES

Favalli, Andrea; Iliev, Metodi; Ianakiev, Kiril; ...

2017-10-09

High-energy delayed γ-ray spectroscopy is a potential technique for directly assaying spent fuel assemblies and achieving the safeguards goal of quantifying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Requirements for the γ-ray detection system, up to ~6 MeV, can be summarized as follows: high efficiency at high γ-ray energies, high energy resolution, good linearity between γ-ray energy and output signal amplitude, ability to operate at very high count rates, and ease of use in industrial environments such as nuclear facilities. High Purity Germanium Detectors (HPGe) are the state of the artmore » and provide excellent energy resolution but are limited in their count rate capability. Lanthanum Bromide (LaBr 3) scintillation detectors offer significantly higher count rate capabilities at lower energy resolution. Thus, LaBr 3 detectors may be an effective alternative for nuclear spent-fuel applications, where count-rate capability is a requirement. This paper documents the measured performance of a 2” (length) × 2” (diameter) of LaBr3 scintillation detector system, coupled to a negatively biased PMT and a tapered active high voltage divider, with count-rates up to ~3 Mcps. An experimental methodology was developed that uses the average current from the PMT’s anode and a dual source method to characterize the detector system at specific very high count rate values. Delayed γ-ray spectra were acquired with the LaBr 3 detector system at the Idaho Accelerator Center, Idaho State University, where samples of ~3g of 235U were irradiated with moderated neutrons from a photo-neutron source. Results of the spectroscopy characterization and analysis of the delayed γ-ray spectra acquired indicate the possible use of LaBr3 scintillation detectors when high count rate capability may outweigh the lower energy resolution.« less

Delayed gamma-ray spectroscopy with lanthanum bromide detector for non-destructive assay of nuclear material

NASA Astrophysics Data System (ADS)

Favalli, Andrea; Iliev, Metodi; Ianakiev, Kiril; Hunt, Alan W.; Ludewigt, Bernhard

2018-01-01

High-energy delayed γ-ray spectroscopy is a potential technique for directly assaying spent fuel assemblies and achieving the safeguards goal of quantifying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Requirements for the γ-ray detection system, up to ∼6 MeV, can be summarized as follows: high efficiency at high γ-ray energies, high energy resolution, good linearity between γ-ray energy and output signal amplitude, ability to operate at very high count rates, and ease of use in industrial environments such as nuclear facilities. High Purity Germanium Detectors (HPGe) are the state of the art and provide excellent energy resolution but are limited in their count rate capability. Lanthanum Bromide (LaBr3) scintillation detectors offer significantly higher count rate capabilities at lower energy resolution. Thus, LaBr3 detectors may be an effective alternative for nuclear spent-fuel applications, where count-rate capability is a requirement. This paper documents the measured performance of a 2" (length) × 2" (diameter) of LaBr3 scintillation detector system, coupled to a negatively biased PMT and a tapered active high voltage divider, with count-rates up to ∼3 Mcps. An experimental methodology was developed that uses the average current from the PMT's anode and a dual source method to characterize the detector system at specific very high count rate values. Delayed γ-ray spectra were acquired with the LaBr3 detector system at the Idaho Accelerator Center, Idaho State University, where samples of ∼3g of 235U were irradiated with moderated neutrons from a photo-neutron source. Results of the spectroscopy characterization and analysis of the delayed γ-ray spectra acquired indicate the possible use of LaBr3 scintillation detectors when high count rate capability may outweigh the lower energy resolution.

New measurement of the 242Pu(n,γ) cross section at n_TOF

NASA Astrophysics Data System (ADS)

Lerendegui-Marco, J.; Guerrero, C.; Cortés-Giraldo, M. A.; Quesada, J. M.; Mendoza, E.; Cano-Ott, D.; Eberhardt, K.; Junghans, A.

2016-03-01

The use of MOX fuel (mixed-oxide fuel made of UO2 and PuO2) in nuclear reactors allows substituting a large fraction of the enriched Uranium by Plutonium reprocessed from spent fuel. With the use of such new fuel composition rich in Pu, a better knowledge of the capture and fission cross sections of the Pu isotopes becomes very important. In particular, a new series of cross section evaluations have been recently carried out jointly by the European (JEFF) and United States (ENDF) nuclear data agencies. For the case of 242Pu, the two only neutron capture time-of-flight measurements available, from 1973 and 1976, are not consistent with each other, which calls for a new time-of flight capture cross section measurement. In order to contribute to a new evaluation, we have perfomed a neutron capture cross section measurement at the n_TOF-EAR1 facility at CERN using four C6D6 detectors, using a high purity target of 95 mg. The preliminary results assessing the quality and limitations (background, statistics and γ-flash effects) of this new experimental data are presented and discussed, taking into account that the aimed accuracy of the measurement ranges between 7% and 12% depending on the neutron energy region.

Status of reduced enrichment programs for research reactors in Japan

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

Kanda, Keiji; Nishihara, Hedeaki; Shirai, Eiji

1997-08-01

The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for themore » full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE.« less

Final Technical Report: Imaging a Dry Storage Cask with Cosmic Ray Muons

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

Yang, Haori; Hayward, Jason; Chichester, David

The goal of this project is to build a scaled prototype system for monitoring used nuclear fuel (UNF) dry storage casks (DSCs) through cosmic ray muon imaging. Such a system will have the capability of verifying the content inside a DSC without opening it. Because of the growth of the nuclear power industry in the U.S. and the policy decision to ban reprocessing of commercial UNF, the used fuel inventory at commercial reactor sites has been increasing. Currently, UNF needs to be moved to independent spent fuel storage installations (ISFSIs), as its inventory approaches the limit on capacity of on-sitemore » wet storage. Thereafter, the fuel will be placed in shipping containers to be transferred to a final disposal site. The ISFSIs were initially licensed as temporary facilities for ~20-yr periods. Given the cancellation of the Yucca mountain project and no clear path forward, extended dry-cask storage (~100 yr.) at ISFSIs is very likely. From the point of view of nuclear material protection, accountability and control technologies (MPACT) campaign, it is important to ensure that special nuclear material (SNM) in UNF is not stolen or diverted from civilian facilities for other use during the extended storage.« less

Uranium mining in France

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

White, L.

Since the onset of the first ''oil shock'' in 1974, France has pursued a policy of steadily increasing energy independence based on nuclear power for generation of electricity. In 1973, nuclear reactors supplied only 8% of France's electrical power. A strong development effort lifted the nuclear share to 23% in 1980, to 66% in 1985, and the plan is to raise the total to 75% by 1990. In 1976, Cogema (Compagnie Generale des Matieres Nucleaires) was organized from the production division of France's Commissariat a l'Energie Atomique (CEA) to handle fuel supply and spent fuel reprocessing for the expanding industrymore » (see subsequent article on Cogema). In parallel with growth of the French nuclear power, Cogema has become a world leader in all aspects of the fuel cycle, providing services not only domestically but internationally as well. As a uranium mining company, Cogema has steadily developed domestic and foreign sources of supply, and over the years it has maintained the world's strongest uranium exploration effort throughout the ups and downs of the market. As a result, the company has become the world's leading uranium supplier, with about 20% of total production contributed either by its domestic mining divisions or overseas subsidiaries.« less

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

NASA Astrophysics Data System (ADS)

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

2010-12-01

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

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

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.

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

Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

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

Ludewigt, Bernhard; Mozin, Vladimir; Campbell, Luke

2015-06-01

High-­energy, beta-delayed gamma-­ray spectroscopy is a potential, non-­destructive assay techniques for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle and for directly assaying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification. Experimental measurements were performed to evaluate fission fragment yields, to test methods for determining isotopic fractions, and to benchmark the modeling code package. Experimental measurement campaigns were carried outmore » at the IAC using a photo-­neutron source and at OSU using a thermal neutron beam from the TRIGA reactor to characterize the emission of high-­energy delayed gamma rays from 235U, 239Pu, and 241Pu targets following neutron induced fission. Data were collected for pure and combined targets for several irradiation/spectroscopy cycle times ranging from 10/10 seconds to 15/30 minutes.The delayed gamma-ray signature of 241Pu, a significant fissile constituent in spent fuel, was measured and compared to 239Pu. The 241Pu/ 239Pu ratios varied between 0.5 and 1.2 for ten prominent lines in the 2700-­3600 keV energy range. Such significant differences in relative peak intensities make it possible to determine relative fractions of these isotopes in a mixed sample. A method for determining fission product yields by fitting the energy and time dependence of the delayed gamma-­ray emission was developed and demonstrated on a limited 235U data set. De-­convolution methods for determining fissile fractions were developed and tested on the experimental data. The use of high count-­rate LaBr 3 detectors was investigated as a potential alternative to HPGe detectors. Modeling capabilities were added to an existing framework and codes were adapted as needed for analyzing experiments and assessing application-­specific assay concepts. A de-­convolution analysis of the delayed gamma-­ray response spectra modeled for spent fuel assemblies was performed using the same method that was applied to the experimental spectra.« less

U.S. sent fuel shipment experience by rail

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

Colborn, K.

2007-07-01

As planning for the large scale shipment of spent nuclear fuel to Yucca Mountain proceeds to address these challenges, actual shipments of spent fuel in other venues continues to provide proof that domestic rail spent fuel shipments can proceed safely and effectively. This paper presents some examples of recently completed spent fuel shipments, and the shipment of large low-level radioactive waste shipments offering lessons learned that may be beneficial to the planning process for large scale spent fuel shipments in the US. (authors)

Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

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

2017-09-01

The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. 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 energy. 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. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

THE ATTRACTIVENESS OF MATERIAS ASSOCIATED WITH THORIUM-BASED NUCLEAR FUEL CYCLES FOR PHWRS

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

Prichard, Andrew W.; Niehus, Mark T.; Collins, Brian A.

2011-07-17

This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with thorium based nuclear fuel cycles. Specifically, this paper examines a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of natural uranium/233U/thorium. This paper uses a PHWR fueled with natural uranium as a base fuel cycle, and then compares material attractiveness of fuel cycles that use 233U/thorium salted with natural uranium. The results include the material attractiveness of fuel at beginning of life (BoL), end of life (EoL), and the number of fuel assemblies requiredmore » to collect a bare critical mass of plutonium or uranium. This study indicates what is required to render the uranium as having low utility for use in nuclear weapons; in addition, this study estimates the increased number of assemblies required to accumulate a bare critical mass of plutonium that has a higher utility for use in nuclear weapons. This approach identifies that some fuel cycles may be easier to implement the International Atomic Energy Agency (IAEA) safeguards approach and have a more effective safeguards by design outcome. For this study, approximately one year of fuel is required to be reprocessed to obtain one bare critical mass of plutonium. Nevertheless, the result of this paper suggests that all spent fuel needs to be rigorously safeguarded and provided with high levels of physical protection. This study was performed at the request of the United States Department of Energy /National Nuclear Security Administration (DOE/NNSA). The methodology and key findings will be presented.« less

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

Argon Shrouded Plasma Spraying of Tantalum over Titanium for Corrosion Protection in Fluorinated Nitric Acid Media

NASA Astrophysics Data System (ADS)

Vetrivendan, E.; Jayaraj, J.; Ningshen, S.; Mallika, C.; Kamachi Mudali, U.

2018-02-01

Argon shrouded plasma spraying (ASPS) was used to deposit a Ta coating on commercially pure Ti (CP-Ti) under inert argon, for dissolver vessel application in the aqueous spent fuels reprocessing plant with high plutonium content. Oxidation during plasma spraying was minimized by shrouding argon system. Porosity and oxide content were controlled by optimizing the spraying parameters, to obtain a uniform and dense Ta coating. The Ta particle temperature and velocity were optimized by judiciously controlling the spray parameters, using a spray diagnostic charge-coupled device camera. The corrosion resistance of the Ta coatings developed by ASPS was investigated by electrochemical studies in 11.5 M HNO3 and 11.5 M HNO3 + 0.05 M NaF. Similarly, the durability of the ASPS Ta coating/substrate was evaluated as per ASTM A262 Practice-C test in boiling nitric acid and fluorinated nitric acid for 240 h. The ASPS Ta coating exhibited higher corrosion resistance than the CP-Ti substrate, as evident from electrochemical studies, and low corrosion rate with excellent coating stability in boiling nitric, and fluorinated nitric acid. The results of the present study revealed that tantalum coating by ASPS is a promising strategy for improving the corrosion resistance in the highly corrosive reprocessing environment.

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

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

NASA Astrophysics Data System (ADS)

Kollar, Lenka

Many countries are planning to build or already are building new nuclear power plants to match their growing energy needs. Since all nuclear power plants handle nuclear materials that could potentially be converted and used for nuclear weapons, they each present a nuclear proliferation risk. Spent nuclear fuel presents the largest build-up of nuclear material at a power plant. This is a proliferation risk because spent fuel contains plutonium that can be chemically separated and used for a nuclear weapon. The International Atomic Energy Agency (IAEA) safeguards spent fuel in all non-nuclear weapons states that are party to the Non-Proliferation Treaty. Various safeguards methods are in use at nuclear power plants and research is underway to develop safeguards methods for spent fuel in centralized storage or underground storage and disposal. Each method of spent fuel storage presents different proliferation risks due to the nature of the storage method and the safeguards techniques that are utilized. Previous proliferation resistance and proliferation risk assessments have mainly compared nuclear material through the whole fuel cycle and not specifically focused on spent fuel storage. This project evaluates the proliferation resistance of the three main types of spent fuel storage: spent fuel pool, dry cask storage, and geological repository. The proliferation resistance assessment methodology that is used in this project is adopted from previous work and altered to be applicable to spent fuel storage. The assessment methodology utilizes various intrinsic and extrinsic proliferation-resistant attributes for each spent fuel storage type. These attributes are used to calculate a total proliferation resistant (PR) value. The maximum PR value is 1.00 and a greater number means that the facility is more proliferation resistant. Current data for spent fuel storage in the United States and around the world was collected. The PR values obtained from this data are 0.49 for the spent fuel pool, 0.42 for dry cask storage, 0.36 for the operating geological repository, and 0.28 for the closed geological repository. Therefore, the spent fuel pool is currently the most proliferation resistant method for storing spent fuel. The extrinsic attributes, mainly involving safeguards measures, affect the total PR value the most. As a result, several recommendations are made to improve the proliferation resistance of spent fuel. These recommendations include employing more advanced safeguards measures, such as verification techniques and remote monitoring, for dry cask storage and the geological repository. Dry cask storage facilities should also be located at the plant and in a secure building to minimize the proliferation risk. Finally, the cost-benefit analysis of increased safeguards needs to be considered. Taking these recommendations into account, the PR values of dry cask storage and the closed geological would be significantly increased, to 0.57 and 0.51, respectively. As a result, with increased safeguards to the safeguards level of the spent fuel pool, dry cask storage would be the most proliferation resistant method to store spent fuel. Therefore, the IAEA should continue to develop remote monitoring and cask storage verification techniques in order to improve the proliferation resistance of spent fuel.

Nuclear energy in Europe: uranium flow modeling and fuel cycle scenario trade-offs from a sustainability perspective.

PubMed

Tendall, Danielle M; Binder, Claudia R

2011-03-15

The European nuclear fuel cycle (covering the EU-27, Switzerland and Ukraine) was modeled using material flow analysis (MFA).The analysis was based on publicly available data from nuclear energy agencies and industries, national trade offices, and nongovernmental organizations. Military uranium was not considered due to lack of accessible data. Nuclear fuel cycle scenarios varying spent fuel reprocessing, depleted uranium re-enrichment, enrichment assays, and use of fast neutron reactors, were established. They were then assessed according to environmental, economic and social criteria such as resource depletion, waste production, chemical and radiation emissions, costs, and proliferation risks. The most preferable scenario in the short term is a combination of reduced tails assay and enrichment grade, allowing a 17.9% reduction of uranium demand without significantly increasing environmental, economic, or social risks. In the long term, fast reactors could theoretically achieve a 99.4% decrease in uranium demand and nuclear waste production. However, this involves important costs and proliferation risks. Increasing material efficiency is not systematically correlated with the reduction of other risks. This suggests that an overall optimization of the nuclear fuel cycle is difficult to obtain. Therefore, criteria must be weighted according to stakeholder interests in order to determine the most sustainable solution. This paper models the flows of uranium and associated materials in Europe, and provides a decision support tool for identifying the trade-offs of the alternative nuclear fuel cycles considered.

The Environmental Protection Agency's Safety Standards for Disposal of Spent Nuclear Fuel: Potential Path Forward in Response to the Report of the Blue Ribbon Commission on America's Nuclear Future - 13388

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

Forinash, Betsy; Schultheisz, Daniel; Peake, Tom

2013-07-01

Following the decision to withdraw the Yucca Mountain license application, the Department of Energy created a Blue Ribbon Commission (BRC) on America's Nuclear Future, tasked with recommending a national strategy to manage the back end of the nuclear fuel cycle. The BRC issued its final report in January 2012, with recommendations covering transportation, storage and disposal of spent nuclear fuel (SNF); potential reprocessing; and supporting institutional measures. The BRC recommendations on disposal of SNF and high-level waste (HLW) are relevant to the U.S. Environmental Protection Agency (EPA), which shares regulatory responsibility with the Nuclear Regulatory Commission (NRC): EPA issues 'generallymore » applicable' performance standards for disposal repositories, which are then implemented in licensing. For disposal, the BRC endorses developing one or more geological repositories, with siting based on an approach that is adaptive, staged and consent-based. The BRC recommends that EPA and NRC work cooperatively to issue generic disposal standards-applying equally to all sites-early in any siting process. EPA previously issued generic disposal standards that apply to all sites other than Yucca Mountain. However, the BRC concluded that the existing regulations should be revisited and revised. The BRC proposes a number of general principles to guide the development of future regulations. EPA continues to review the BRC report and to assess the implications for Agency action, including potential regulatory issues and considerations if EPA develops new or revised generic disposal standards. This review also involves preparatory activities to define potential process and public engagement approaches. (authors)« less

An optimization methodology for heterogeneous minor actinides transmutation

NASA Astrophysics Data System (ADS)

Kooyman, Timothée; Buiron, Laurent; Rimpault, Gérald

2018-04-01

In the case of a closed fuel cycle, minor actinides transmutation can lead to a strong reduction in spent fuel radiotoxicity and decay heat. In the heterogeneous approach, minor actinides are loaded in dedicated targets located at the core periphery so that long-lived minor actinides undergo fission and are turned in shorter-lived fission products. However, such targets require a specific design process due to high helium production in the fuel, high flux gradient at the core periphery and low power production. Additionally, the targets are generally manufactured with a high content in minor actinides in order to compensate for the low flux level at the core periphery. This leads to negative impacts on the fuel cycle in terms of neutron source and decay heat of the irradiated targets, which penalize their handling and reprocessing. In this paper, a simplified methodology for the design of targets is coupled with a method for the optimization of transmutation which takes into account both transmutation performances and fuel cycle impacts. The uncertainties and performances of this methodology are evaluated and shown to be sufficient to carry out scoping studies. An illustration is then made by considering the use of moderating material in the targets, which has a positive impact on the minor actinides consumption but a negative impact both on fuel cycle constraints (higher decay heat and neutron) and on assembly design (higher helium production and lower fuel volume fraction). It is shown that the use of moderating material is an optimal solution of the transmutation problem with regards to consumption and fuel cycle impacts, even when taking geometrical design considerations into account.

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.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-13

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

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

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

Principles of Product Quality Control of German Radioactive Waste Forms from the Reprocessing of Spent Fuel: Vitrification, Compaction and Numerical Simulation - 12529

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

Tietze-Jaensch, Holger; Schneider, Stephan; Aksyutina, Yuliya

2012-07-01

The German product quality control is inter alia responsible for control of two radioactive waste forms of heat generating waste: a) homogeneous vitrified HLW and b) heterogeneous compacted hulls, end-pieces and technological metallic waste. In either case, significantly different metrology is employed at the site of the conditioning plant for the obligatory nuclide inventory declaration. To facilitate an independent evaluation and checking of the accompanying documentation numerical simulations are carried out. The physical and chemical properties of radioactive waste residues are used to assess the data consistency and uncertainty margins, as well as to predict the long-term behavior of themore » radioactive waste. This is relevant for repository acceptance and safety considerations. Our new numerical approach follows a bottom-up simulation starting from the burn-up behavior of the fuel elements in the reactor core. The output of these burn-up calculations is then coupled with a program that simulates the material separation in the subsequent dissolution and extraction processes normalized to the mass balance. Follow-up simulations of the separated reprocessing lines of a) the vitrification of highly-active liquid and b) the compaction of residual intermediate-active metallic hulls remaining after fuel pellets dissolution, end-pieces and technological waste, allows calculating expectation values for the various repository relevant properties of either waste stream. The principles of the German product quality control of radioactive waste residues from the spent fuel reprocessing have been introduced and explained. Namely, heat generating homogeneous vitrified HLW and heterogeneous compacted metallic MLW have been discussed. The advantages of a complementary numerical property simulation have been made clear and examples of benefits are presented. We have compiled a new program suite to calculate the physical and radio-chemical properties of common nuclear waste residues. The immediate benefit is the independent assessment of radio-active inventory declarations and much facilitated product quality control of waste residues that need to be returned to Germany and submitted to a German HLW-repository requirements. Wherever possible, internationally accepted standard programs are used and embedded. The innovative coupling of burn-up calculations (SCALE) with neutron and gamma transport codes (MCPN-X) allows an application in the world of virtual waste properties. If-then-else scenarios of hypothetical waste material compositions and distributions provide valuable information of long term nuclide property propagation under repository conditions over a very long time span. Benchmarking the program with real residue data demonstrates the power and remarkable accuracy of this numerical approach, boosting the reliability of the confidence aforementioned numerous applications, namely the proof tool set for on-the-spot production quality checking and data evaluation and independent verification. Moreover, using the numerical bottom-up approach helps to avoid the accumulation of fake activities that may gradually build up in a repository from the so-called conservative or penalizing nuclide inventory declarations. The radioactive waste properties and the hydrolytic and chemical stability can be predicted. The interaction with invasive chemicals can be assessed and propagation scenarios can be developed from reliable and sound data and HLW properties. Hence, the appropriate design of a future HLW repository can be based upon predictable and quality assured waste characteristics. (authors)« less

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

Code of Federal Regulations, 2010 CFR

2010-01-01

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

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

Code of Federal Regulations, 2011 CFR

2011-01-01

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

Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

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

Pond, R.B.; Matos, J.E.

1996-05-01

As part of the Department of Energy`s spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, aremore » not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report.« less

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

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

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

Shinichi Aose; Takafumi Kitajima; Kouji Ogasawara

CPF (Chemical Processing Facility) was constructed at Nuclear Fuel Cycle Engineering Laboratories of JAEA (Japan Atomic Energy Agency) in 1980 as a basic research field where spent fuel pins from fast reactor (FR) and high level liquid waste can be dealt with. The renovation consists of remodeling of the CA-3 cell and the laboratory A, installation of globe boxes, hoods and analytical equipments to the laboratory C and the analytical laboratory. Also maintenance equipments in the CA-5 cell which had been out of order were repaired. The CA-3 cell is the main cell in which important equipments such as amore » dissolver, a clarifier and extractors are installed for carrying out the hot test using the irradiated FR fuel. Since the CPF had specialized originally in the research function for the Purex process, it was desired to execute the research and development of such new, various reprocessing processes. Formerly, equipments were arranged in wide space and connected with not only each other but also with utility supply system mainly by fixed stainless steel pipes. It caused shortage of operation space in flexibility for basic experimental study. Old equipments in the CA-3 cell including vessels and pipes were removed after successful decontamination, and new equipments were installed conformably to the new design. For the purpose of easy installation and rearranging the experimental equipments, equipments are basically connected by flexible pipes. Since dissolver is able to be easily replaced, various dissolution experiments is conducted. Insoluble residue generated by dissolution of spent fuel is clarified by centrifugal. This small apparatus is effective to space-saving. Mini mixer settlers or centrifugal contactors are put on to the prescribed limited space in front of the backside wall. Fresh reagents such as solvent, scrubbing and stripping solution are continuously fed from the laboratory A to the extractor by the reagent supply system with semi-automatic observation system. The in-cell crane in CA-5 was renovated to increase driving efficiency. At the renovation for the in-cell crane, full scale mockup test and 3D simulation test had been executed in advance. After the renovation, hot tests in the CPF had been resumed from JFY 2002. New equipments such as dissolver, extractor, electrolytic device, etc. were installed in CA-3 conformably to the new design laid out in order to ensure the function and space. Glove boxes in the analysis laboratory were renewed in order to let it have flexibility from the viewpoint of conducting basic experiments (ex. U crystallization). Glove boxes and hoods were newly installed in the laboratory A for basic research and analysis, especially on MA chemistries. One laboratory (the laboratory C) was established to research about dry reprocessing. The renovation of the CPF has been executed in order to contribute to the development on the advanced fast reactor fuel cycle system, which will give us many sort of technical subject and experimental theme to be solved in the 2. Generation of the CPF.« less

SCAN+

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

Kenneth Krebs, John Svoboda

2009-11-01

SCAN+ is a software application specifically designed to control the positioning of a gamma spectrometer by a two dimensional translation system above spent fuel bundles located in a sealed spent fuel cask. The gamma spectrometer collects gamma spectrum information for the purpose of spent fuel cask fuel loading verification. SCAN+ performs manual and automatic gamma spectrometer positioning functions as-well-as exercising control of the gamma spectrometer data acquisitioning functions. Cask configuration files are used to determine the positions of spent fuel bundles. Cask scanning files are used to determine the desired scan paths for scanning a spent fuel cask allowing formore » automatic unattended cask scanning that may take several hours.« less

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

Code of Federal Regulations, 2011 CFR

2011-01-01

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

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

Code of Federal Regulations, 2010 CFR

2010-01-01

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

Sodium aluminum-iron phosphate glass-ceramics for immobilization of lanthanide oxide wastes from pyrochemical reprocessing of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Stefanovsky, S. V.; Stefanovsky, O. I.; Kadyko, M. I.; Nikonov, B. S.

2018-03-01

Sodium aluminum (iron) phosphate glass ceramics containing of up to 20 wt.% rare earth (RE) oxides simulating pyroprocessing waste were produced by melting at 1250 °C followed by either quenching or slow cooling to room temperature. The iron-free glass-ceramics were composed of major glass and minor phosphotridymite and monazite. The iron-bearing glass-ceramics were composed of major glass and minor monazite and Na-Al-Fe orthophosphate at low waste loadings (5-10 wt.%) and major orthophosphate and minor monazite as well as interstitial glass at high waste loadings (15-20 wt.%). Slowly cooled samples contained higher amount of crystalline phases than quenched ones. Monazite is major phase for REs. Leach rates from the materials of major elements (Na, Al, Fe, P) are 10-5-10-7 g cm-2 d-1, RE elements - lower than 10-5 g cm-2 d-1.

Corrosion behavior of Alloy 690 and Alloy 693 in simulated nuclear high level waste medium

NASA Astrophysics Data System (ADS)

Samantaroy, Pradeep Kumar; Suresh, Girija; Paul, Ranita; Kamachi Mudali, U.; Raj, Baldev

2011-11-01

Nickel based alloys are candidate materials for the storage of high level waste (HLW) generated from reprocessing of spent nuclear fuel. In the present investigation Alloy 690 and Alloy 693 are assessed by potentiodynamic anodic polarization technique for their corrosion behavior in 3 M HNO 3, 3 M HNO 3 containing simulated HLW and in chloride medium. Both the alloys were found to possess good corrosion resistance in both the media at ambient condition. Microstructural examination was carried out by SEM for both the alloys after electrolytic etching. Compositional analysis of the passive film formed on the alloys in 3 M HNO 3 and 3 M HNO 3 with HLW was carried out by XPS. The surface of Alloy 690 and Alloy 693, both consists of a thin layer of oxide of Ni, Cr, and Fe under passivation in both the media. The results of investigation are presented in the paper.

Modeling the potential radionuclide transport by the Ob and Yenisey Rivers to the Kara Sea.

PubMed

Paluszkiewicz, T; Hibler, L F; Richmond, M C; Bradley, D J; Thomas, S A

2001-01-01

A major portion of the former Soviet Union (FSU) nuclear program is located in the West Siberian Basin. Among the many nuclear facilities are three production reactors and the spent nuclear fuel reprocessing sites, Mayak, Tomsk-7, and Krasnoyarsk-26, which together are probably responsible for the majority of the radioactive contamination found in the Ob and Yenisey River systems that feed into the Arctic Ocean through the Kara Sea. This manuscript describes ongoing research to estimate radionuclide fluxes to the Kara Sea from these river systems. Our approach is to apply a hierarchy of simple models that use existing and forthcoming data to quantify the transport and fate of radionuclide contaminants via various environmental pathways. We present an initial quantification of the contaminant inventory, hydrology, meteorology, and sedimentology of the Ob River system and preliminary conclusions from portions of the Ob River model.

The Nuclear Energy Advanced Modeling and Simulation Safeguards and Separations Reprocessing Plant Toolkit

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

McCaskey, Alex; Billings, Jay Jay; de Almeida, Valmor F

2011-08-01

This report details the progress made in the development of the Reprocessing Plant Toolkit (RPTk) for the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. RPTk is an ongoing development effort intended to provide users with an extensible, integrated, and scalable software framework for the modeling and simulation of spent nuclear fuel reprocessing plants by enabling the insertion and coupling of user-developed physicochemical modules of variable fidelity. The NEAMS Safeguards and Separations IPSC (SafeSeps) and the Enabling Computational Technologies (ECT) supporting program element have partnered to release an initial version of the RPTk with a focus on software usabilitymore » and utility. RPTk implements a data flow architecture that is the source of the system's extensibility and scalability. Data flows through physicochemical modules sequentially, with each module importing data, evolving it, and exporting the updated data to the next downstream module. This is accomplished through various architectural abstractions designed to give RPTk true plug-and-play capabilities. A simple application of this architecture, as well as RPTk data flow and evolution, is demonstrated in Section 6 with an application consisting of two coupled physicochemical modules. The remaining sections describe this ongoing work in full, from system vision and design inception to full implementation. Section 3 describes the relevant software development processes used by the RPTk development team. These processes allow the team to manage system complexity and ensure stakeholder satisfaction. This section also details the work done on the RPTk ``black box'' and ``white box'' models, with a special focus on the separation of concerns between the RPTk user interface and application runtime. Section 4 and 5 discuss that application runtime component in more detail, and describe the dependencies, behavior, and rigorous testing of its constituent components.« less

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

Spent Nuclear Fuel

EIA Publications

2015-01-01

Spent nuclear fuel data are collected by the U.S. Energy Information Administration (EIA) for the Department of Energy's Office of Standard Contract Management (Office of the General Counsel) on the Form GC-859, "Nuclear Fuel Data Survey." The data include detailed characteristics of spent nuclear fuel discharged from commercial U.S. nuclear power plants and currently stored at commercial sites in the United States. Utilities were not required to report spent nuclear fuel assemblies shipped to away-from-reactor, off-site facilities.

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

Spent nuclear fuel dry transfer system

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

Stewart, L.; Agace, S.

The U.S. Department of Energy is currently engaged in a cooperative program with the Electric Power Research Institute (EPRI) to design a spent nuclear fuel dry transfer system (DTS). The system will enable the transfer of individual spent nuclear fuel assemblies between a conventional top loading cask and multi-purpose canister in a shielded overpack, or accommodate spent nuclear fuel transfers between two conventional casks.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-07

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

77 FR 76952 - Rescinding Spent Fuel Pool Exclusion Regulations

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-31

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

Annual report, FY 1979 Spent fuel and fuel pool component integrity.

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

Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.

International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-..mu..m) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion.more » A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report.« 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.

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-29

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

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-03

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

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

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

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

1995-12-01

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

Recovery of minor actinides from spent fuel using TPEN-immobilized gels

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

Koyama, S.; Suto, M.; Ohbayashi, H.

2013-07-01

A series of separation experiments was performed in order to study the recovery process for minor actinides (MAs), such as americium (Am) and curium (Cm), from the actual spent fuel by using an extraction chromatographic technique. N,N,N',N'-tetrakis-(4-propenyloxy-2-pyridylmethyl) ethylenediamine (TPPEN) is an N,N,N',N'-tetrakis (2-pyridylmethyl) ethylenediamine (TPEN) analogue consisting of an incorporated pyridine ring that acts as not only a ligand but also as a site for polymerization and crosslinking of the gel. The TPPEN and N-isopropylacrylamide (NIPA) were dissolved into dimethylformamide (DMF, Wako Co., Ltd.) and a silica beads polymer, and then TTPEN was immobilized chemically in a polymer gel (somore » called TPEN-gel). Mixed oxide (MOX) fuel, which was highly irradiated up to 119 GWD/MTM in the experimental fast reactor Joyo, was used as a reference spent fuel. First, uranium (U) and plutonium (Pu) were separated from the irradiated fuel using an ion-exchange method, and then, the platinum group elements were removed by CMPO to leave a mixed solution of MAs and lanthanides. The 3 mol% TPPEN-gel was packed with as an extraction column (CV: 1 ml) and then rinsed by 0.1 M NaNO{sub 3}(pH 4.0) for pH adjustment. After washing the column by 0.01 M NaNO{sub 3} (pH 4.0), Eu was detected and the recovery rate reached 93%. The MAs were then recovered by changing the eluent to 0.01 M NaNO{sub 3} (pH 2.0), and the recovery rate of Am was 48 %. The 10 mol% TPPEN-gel was used to improve adsorption coefficient of Am and a condition of eluent temperature was changed in order to confirm the temperature swing effect on TPEN-gel for MA. More than 90% Eu was detected in the eluent after washing with 0.01 M NaNO{sub 3} (pH 3.5) at 5 Celsius degrees. Americium was backwardly detected and eluted continuously during the same condition. After removal of Eu, the eluent temperature was changed to 32 Celsius degrees, then Am was detected (pH 3.0). Finally remained Am could be stripped from TPPEN-gel by changing the pH of the eluent to 2.0. These results These results prove that the proposed recovery process for MAs is a potential candidate for future reprocessing methods based on the extraction chromatographic technique. (authors)« 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

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

Transportation accident scenarios for commercial spent fuel

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

Wilmot, E L

1981-02-01

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

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

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

Ohta, Hirokazu; Inoue, Tadashi; Sakamura, Yoshiharu

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

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

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.

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.

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

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

Code of Federal Regulations, 2011 CFR

2011-01-01

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

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

Code of Federal Regulations, 2010 CFR

2010-01-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-13

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

Impact of the deployment schedule of fast breeding reactors in the frame of French act for nuclear materials and radioactive waste management

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

Le Mer, J.; Garzenne, C.; Lemasson, D.

In the frame of the French Act of June 28, 2006 on 'a sustainable management of nuclear materials and radioactive waste' EDF R and D assesses various research scenarios of transition between the actual French fleet and a Generation IV fleet with a closed fuel cycle where plutonium is multi-recycled. The basic scenarios simulate a deployment of 60 GWe of Sodium-cooled Fast Reactors (SFRs) in two steps: one third from 2040 to 2050 and the rest from 2080 to 2100 (scenarios 2040). These research scenarios assume that SFR technology will be ready for industrial deployment in 2040. One of themore » many sensitivity analyses that EDF, as a nuclear power plant operator, must evaluate is the impact of a delay of SFR technology in terms of uranium consumptions, plutonium needs and fuel cycle utilities gauging. The sensitivity scenarios use the same assumptions as scenarios 2040 but they simulate a different transition phase: SFRs are deployed in one step between 2080 and 2110 (scenarios 2080). As the French Act states to conduct research on minor actinides (MA) management, we studied different options for 2040 and 2080 scenarios: no MA transmutation, americium transmutation in heterogeneous mode based on americium Bearing Blankets (AmBB) in SFRs and all MA transmutation in heterogeneous mode based on MA Bearing Blankets (MABB). Moreover, we studied multiple parameters that could impact the deployment of these reactors (SFR load factor, increase of the use of MOX in Light Water Reactors, increase of the cooling time in spent nuclear fuel storage...). Each scenario has been computed with the EDF R and D fuel cycle simulation code TIRELIRE-STRATEGIE and optimized to meet various fuel cycle constraints such as using the reprocessing facility with long period of constant capacity, keeping the temporary stored mass of plutonium and MA under imposed limits, recycling older assemblies first... These research scenarios show that the transition from the current PWR fleet to an equivalent fleet of Generation IV SFR can follow different courses. The design of SFR-V2B that we used in our studies needs a high inventory of plutonium resulting in tension on this resource. Several options can be used in order to loosen this tension: our results lead to favour the use of axial breeding blanket in SFR. Load factor of upcoming reactors has to be regarded with attention as it has a high impact on plutonium resource for a given production of electricity. The deployment of SFRs beginning in 2080 instead of 2040 following the scenarios we described creates higher tensions on reprocessing capacity, separated plutonium storage and spent fuel storage. In the frame of the French Act, we studied minor actinides transmutation. The flux of MA in all fuel cycle plants is really high, which will lead to decay heat, a and neutron emission related problems. In terms of reduction of MA inventories, the deployment of MA transmutation cycle must not delay the installation of SFRs. The plutonium production in MABB and AmBB does not allow reducing the use of axial breeding blankets. The impact of MA or Am transmutation over the high level waste disposal is more important if the SFRs are deployed later. Transmutation option (americium or all MA) does not have a significant impact on the number of canister produced nor on its long-term thermal properties. (authors)« less

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

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

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

1997-06-01

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

A review and overview of nuclear waste management

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

Murray, R.L.

1984-12-31

An understanding of the status and issues in the management of radioactive wastes is based on technical information on radioactivity, radiation, biological hazard of radiation exposure, radiation standards, and methods of protection. The fission process gives rise to radioactive fission products and neutron bombardment gives activation products. Radioactive wastes are classified according to source: defense, commercial, industrial, and institutional; and according to physical features: uranium mill tailings, high-level, transuranic, and low-level. The nuclear fuel cycle, which contributes a large fraction of annual radioactive waste, starts with uranium ore, includes nuclear reactor use for electrical power generation, and ends with ultimatemore » disposal of residues. The relation of spent fuel storage and reprocessing is governed by technical, economic, and political considerations. Waste has been successfully solidified in glass and other forms and choices of the containers for the waste form are available. Methods of disposal of high-level waste that have been investigated are transmutation by neutron bombardment, shipment to Antartica, deep-hole insertion, subseabed placement, transfer by rocket to an orbit in space, and disposal in a mined cavity. The latter is the favored method. The choices of host geological media are salt, basalt, tuff, and granite.« 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

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

NASA Astrophysics Data System (ADS)

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

2013-07-01

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

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.

The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

DOE PAGES

Bonano, Evaristo J.; Kalinina, Elena A.; Swift, Peter N.

2018-02-26

Current practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-centurymore » when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.« less

The Need for Integrating the Back End of the Nuclear Fuel Cycle in the United States of America

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

Bonano, Evaristo J.; Kalinina, Elena A.; Swift, Peter N.

Current practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-centurymore » when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.« 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

2013-10-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-26

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

Integral nuclear data validation using experimental spent nuclear fuel compositions

DOE PAGES

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

2017-07-19

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

Integral nuclear data validation using experimental spent nuclear fuel compositions

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

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

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

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

DOE PAGES

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

2017-04-28

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

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

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

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

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

Method for producing hydrocarbon fuels and fuel gas from heavy polynuclear hydrocarbons by the use of molten metal halide catalysts

DOEpatents

Gorin, Everett

1979-01-01

In a process for hydrocracking heavy polynuclear carbonaceous feedstocks to produce lighter hydrocarbon fuels by contacting the heavy feedstocks with hydrogen in the presence of a molten metal halide catalyst in a hydrocracking zone, thereafter separating at least a major portion of the lighter hydrocarbon fuels from the spent molten metal halide and thereafter regenerating the spent molten metal halide by incinerating the spent molten metal halide by combustion of carbon and sulfur compounds in the spent molten metal halide in an incineration zone, the improvement comprising: (a) contacting the heavy feedstocks and hydrogen in the presence of the molten metal halide in the hydrocracking zone at reaction conditions effective to convert from about 60 to about 90 weight percent of the feedstock to lighter hydrocarbon fuels; (b) separating at least a major portion of the lighter hydrocarbon fuels from the spent molten metal halide; (c) contacting the spent molten metal halide with oxygen in a liquid phase gasification zone at a temperature and pressure sufficient to vaporize from about 25 to about 75 weight percent of the spent metal halide, the oxygen being introduced in an amount sufficient to remove from about 60 to about 90 weight percent of the carbon contained in the spent molten metal halide to produce a fuel gas and regenerated metal halide; and (d) incinerating the spent molten metal halide by combusting carbon and sulfur compounds contained therein.

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

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

Christopher Landers; Igor Bolshinsky; Ken Allen

2010-07-01

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

10 CFR 961.2 - Applicability.

Code of Federal Regulations, 2010 CFR

2010-01-01

... ENERGY STANDARD CONTRACT FOR DISPOSAL OF SPENT NUCLEAR FUEL AND/OR HIGH-LEVEL RADIOACTIVE WASTE General... owns or generates spent nuclear fuel or high-level radioactive waste, of domestic origin, generated in... part will commit DOE to accept title to, transport, and dispose of such spent fuel and waste. In...

An approach to meeting the spent fuel standard

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

Makhijani, A.

1996-05-01

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

Spent Nuclear Fuel (SNF) Project Execution Plan

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

LEROY, P.G.

2000-11-03

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

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

10 CFR 961.3 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... ENERGY STANDARD CONTRACT FOR DISPOSAL OF SPENT NUCLEAR FUEL AND/OR HIGH-LEVEL RADIOACTIVE WASTE General... means any person who has title to spent nuclear fuel or high-level radioactive waste. Purchaser means... (42 U.S.C. 2133, 2134) or who has title to spent nuclear fuel or high level radioactive waste and who...

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

Code of Federal Regulations, 2010 CFR

2010-01-01

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

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

Code of Federal Regulations, 2010 CFR

2010-01-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-03

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-02

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-25

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-18

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

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

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

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

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

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

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

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

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

New developments and prospects on COSI, the simulation software for fuel cycle analysis

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

Eschbach, R.; Meyer, M.; Coquelet-Pascal, C.

2013-07-01

COSI, software developed by the Nuclear Energy Direction of the CEA, is a code simulating a pool of nuclear power plants with its associated fuel cycle facilities. This code has been designed to study various short, medium and long term options for the introduction of various types of nuclear reactors and for the use of associated nuclear materials. In the frame of the French Act for waste management, scenario studies are carried out with COSI, to compare different options of evolution of the French reactor fleet and options of partitioning and transmutation of plutonium and minor actinides. Those studies aimmore » in particular at evaluating the sustainability of Sodium cooled Fast Reactors (SFR) deployment and the possibility to transmute minor actinides. The COSI6 version is a completely renewed software released in 2006. COSI6 is now coupled with the last version of CESAR (CESAR5.3 based on JEFF3.1.1 nuclear data) allowing the calculations on irradiated fuel with 200 fission products and 100 heavy nuclides. A new release is planned in 2013, including in particular the coupling with a recommended database of reactors. An exercise of validation of COSI6, carried out on the French PWR historic nuclear fleet, has been performed. During this exercise quantities like cumulative natural uranium consumption, or cumulative depleted uranium, or UOX/MOX spent fuel storage, or stocks of reprocessed uranium, or plutonium content in fresh MOX fuel, or the annual production of high level waste, have been computed by COSI6 and compared to industrial data. The results have allowed us to validate the essential phases of the fuel cycle computation, and reinforces the credibility of the results provided by the code.« less

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

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

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

DOE PAGES

Hu, Jianwei; Gauld, Ian C.

2014-12-01

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

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

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

Hu, Jianwei; Gauld, Ian C.

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

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

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

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

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

1991-12-01

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

GEM*STAR: Time for an Alternative Way Forward

NASA Astrophysics Data System (ADS)

Vogelaar, R. Bruce

2011-10-01

The presumption that nuclear reactors will retain their role in global energy production is constantly being challenged - even more so following recent events at Fukushima. Nuclear energy, despite being ``green,'' has inexorably been coupled in the public mind with three paramount concerns: safety, weapons proliferation, and waste (and then ultimately cost). Over the past four decades, the safety of deployed fleets has greatly improved, yet the capital and political costs of a ``nuclear energy option'' appear insurmountable in several countries. The US approach to civilian nuclear energy has become deeply entrenched, first through choices made by the military, and then by the deployed nuclear reactor fleet. This extends to the research agencies as well, to the point where basic sciences and nuclear energy operate in separate spheres. But technologies and priorities have changed, and the time has arrived where a transformative re-think of nuclear energy is not only possible, but urgent. And nuclear physicists are uniquely positioned to accomplish this. This talk will show that by asking, and answering,``what would an accelerator-driven civilian nuclear energy program look like,'' ADNA Corporation's GEM*STAR design directly addresses all three fundamental concerns: safety, proliferation, and waste - and also the final hurdle: cost. GEM*STAR is not an ``add-on'' (to either Project-X, or GEN III+), but rather a base-line energy production capacity, for either electricity or transport fuel production. It integrates and advances the molten-salt reactor technology developed at ORNL, the MW beam accelerator technologies developed by basic sciences, and a reactor/target design optimized for accelerator driven-systems. The results include: the ability to use LWR spent fuel without reprocessing or additional waste; the ability to use natural uranium; no critical mass ever present; orders-of-magnitude less volatile radioactivity in the core; more efficient use of, and deeper burn of actinides, without additional waste; proliferation resistance (no enrichment or reprocessing); high-tolerance to ``beam-trips'' and ultimately, and perhaps most importantly, lower cost electricity or diesel fuel than any currently envisioned new energy source.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-03

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

10 CFR 72.8 - Denial of licensing by Agreement States.

Code of Federal Regulations, 2010 CFR

2010-01-01

... SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General... the storage of spent fuel and reactor-related GTCC waste in an ISFSI or the storage of spent fuel, high-level radioactive waste, and reactor-related GTCC waste in an MRS. [66 FR 51839, Oct. 11, 2001] ...

10 CFR 72.22 - Contents of application: General and financial information.

Code of Federal Regulations, 2011 CFR

2011-01-01

... INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN... of spent fuel, high-level radioactive waste, and/or reactor-related GTCC waste from storage. (f) Each applicant for a license under this part to receive, transfer, and possess power reactor spent fuel, power...

10 CFR 72.8 - Denial of licensing by Agreement States.

Code of Federal Regulations, 2011 CFR

2011-01-01

... SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General... the storage of spent fuel and reactor-related GTCC waste in an ISFSI or the storage of spent fuel, high-level radioactive waste, and reactor-related GTCC waste in an MRS. [66 FR 51839, Oct. 11, 2001] ...

76 FR 54808 - Agency Information Collection Activities: Submission for the Office of Management and Budget (OMB...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-02

... the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste and Reactor-Related... receive, transfer, package and possess power reactor spent fuel, high-level waste, and other radioactive..., package, and possess power reactor spent fuel and high-level radioactive waste, and other associated...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-29

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

Timely topics on spent fuel storage

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

Selin, I.

1994-12-31

The history of spent fuel management in this country has taken several turns, with a final resolution still out of reach. Several repository programs started, stalled ans stopped. The latest effort at Yucca Mountain is progressing but, at best, is years from the early phases of licensing, much less the actual underground disposal of spent fuel. A monitored retrieval storage [MRS] facility was expected to start accepting commercial spent fuel beginning in 1998, but no such facility is clearly on the horizon. All of these recent developments changed the circumstances that we face in spent fuel management. The obvious conclusionmore » is that an increasing number of plants, both operating and permanently shut-down reactors, will have to provide for additional spent fuel storage on-site for a longer period than originally planned, and even after plant decommissioning, prudence requires that provision be made for continual, stand-alone, on-site storage. After pool capacity is reached, most utilities opt for some sort of dry storage. But the dry storage option has triggered an unprecedented amount of local opposition at many sites, further taxing NRC and industry resources.« less

Method for shearing spent nuclear fuel assemblies

DOEpatents

Weil, Bradley S.; Watson, Clyde D.

1977-01-01

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

Spent Fuel Ratio Estimates from Numerical Models in ALE3D

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

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

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

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

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

Field, D.; McAtamney, N.

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

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

Iseki, Tadahiro; Inaba, Makoto; Takahashi, Naoki

During the second and third steps of Active Test at Rokkasho Reprocessing Plant (RRP), the performances of the Separation Facility have been checked; (A) diluent washing efficiency, (B) plutonium stripping efficiency, (C) decontamination factor of fission products and (D) plutonium and uranium leakage into raffinate and spent solvent. Test results were equivalent to or better than expected. (authors)

In-Situ Characterization of Underwater Radioactive Sludge

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

Simpson, A.P.; Clapham, M.J.; Swinson, B.

2008-07-01

A fundamental requirement underpinning safe clean-up technologies for legacy spent nuclear fuel (SNF) ponds, pools and wet silos is the ability to characterize the radioactive waste form prior to retrieval. The corrosion products resulting from the long term underwater storage of spent nuclear fuel, reactor components and reprocessing debris present a major hazard to facility decontamination and decommissioning in terms of their radioactive content and physical / chemical reactivity. The ability to perform in-situ underwater non-destructive characterization of sludge and debris in a safe and cost-effective manner offers significant benefits over traditional destructive sampling methods. Several techniques are available formore » underwater measurements including (i) Gross gamma counting, (ii) Low-, Medium- and High- Resolution Gamma Spectroscopy, (iii) Passive neutron counting and (iv) Active Neutron Interrogation. The optimum technique depends on (i) the radioactive inventory (ii) mechanical access restrictions for deployment of the detection equipment, interrogation sources etc. (iii) the integrity of plant records and (iv) the extent to which Acceptable Knowledge which may be used for 'fingerprinting' the radioactive contents to a marker nuclide. Prior deployments of underwater SNF characterization equipment around the world have been reviewed with respect to recent developments in gamma and neutron detection technologies, digital electronics advancements, data transfer techniques, remote operation capabilities and improved field ruggedization. Modeling and experimental work has been performed to determine the capabilities, performance envelope and operational limitations of the future generation of non-destructive underwater sludge characterization techniques. Recommendations are given on the optimal design of systems and procedures to provide an acceptable level of confidence in the characterization of residual sludge content of legacy wet storage facilities such that retrieval and repackaging of SNF sludges may proceed safely and efficiently with support of the regulators and the public. (author)« less

10 CFR 50.68 - Criticality accident requirements.

Code of Federal Regulations, 2012 CFR

2012-01-01

... boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel... flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel...

10 CFR 50.68 - Criticality accident requirements.

Code of Federal Regulations, 2014 CFR

2014-01-01

... boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel... flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel...

10 CFR 50.68 - Criticality accident requirements.

Code of Federal Regulations, 2013 CFR

2013-01-01

... boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel... flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel...

10 CFR 50.68 - Criticality accident requirements.

Code of Federal Regulations, 2011 CFR

2011-01-01

... boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel... flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel...

10 CFR 50.68 - Criticality accident requirements.

Code of Federal Regulations, 2010 CFR

2010-01-01

... boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel... flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel...

Nondestructive Assay Data Integration with the SKB-50 Assemblies - FY16 Update

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

Tobin, Stephen Joseph; Fugate, Michael Lynn; Trellue, Holly Renee

2016-10-28

A project to research the application of non-destructive assay (NDA) techniques for spent fuel assemblies is underway at the Central Interim Storage Facility for Spent Nuclear Fuel (for which the Swedish acronym is Clab) in Oskarshamn, Sweden. The research goals of this project contain both safeguards and non-safeguards interests. These nondestructive assay (NDA) technologies are designed to strengthen the technical toolkit of safeguard inspectors and others to determine the following technical goals more accurately; Verify initial enrichment, burnup, and cooling time of facility declaration for spent fuel assemblies; Detect replaced or missing pins from a given spent fuel assembly tomore » confirm its integrity; and Estimate plutonium mass and related plutonium and uranium fissile mass parameters in spent fuel assemblies. Estimate heat content, and measure reactivity (multiplication).« less

Spent fuel burnup estimation by Cerenkov glow intensity measurement

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

Kuribara, Masayuki

1994-10-01

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

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

DOE PAGES

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

2017-09-01

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

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

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

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

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

Transuranic contamination of stainless steel in nitric acid

NASA Astrophysics Data System (ADS)

Kerry, Timothy; Banford, Anthony W.; Thompson, Olivia R.; Carey, Thomas; Schild, Dieter; Geist, Andreas; Sharrad, Clint A.

2017-09-01

Stainless steels coupons have been exposed to transuranic species in conditions representative of those found in a spent nuclear fuel reprocessing plant. Stainless steel was prepared to different surface finishes and exposed to nitric acid of varying concentrations containing 237Np, 239Pu or 243Am for one month at 50 °C. Contamination by these transuranics has been observed on all surfaces exposed to the solution through the use of autoradiography. This technique showed that samples held in 4 M HNO3 bind 2-3 times as much radionuclide as those held in 10.5 M HNO3. It was also found that the polished steel surfaces generally took up more transuranic contamination than the etched and "as received" steel finishes. The extent of corrosion on the steel surfaces was found, by scanning electron microscopy, to be greater in solutions containing Np and Pu in comparison to that observed from contact with Am containing solutions, indicating that redox activity of transuranics can influence the mechanism of stainless steel corrosion.

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

Nan, Yue; Tavlarides, Lawrence L.; DePaoli, David W.

The adsorption process of iodine, a major volatile radionuclide in the off-gas streams of spent nuclear fuel reprocessing, on hydrogen-reduced silver-exchanged mordenite (Ag 0Z) was studied at the micro-scale. The gas-solid mass transfer and reaction involved in the adsorption process were investigated and evaluated with appropriate models. Optimal conditions for reducing the silver-exchanged mordenite (AgZ) in a hydrogen stream were determined. Kinetic and equilibrium data of iodine adsorption on Ag 0Z were obtained by performing single-layer adsorption experiments with experimental systems of high precision at 373–473 K over various iodine concentrations. Results indicate approximately 91% to 97% of the iodinemore » adsorption was through the silver-iodine reaction. The effect of temperature on the iodine loading capacity of Ag 0Z was discussed. In conclusion, the Shrinking Core model describes the data well, and the primary rate controlling mechanisms were macro-pore diffusion and silver-iodine reaction. © 2016 American Institute of Chemical Engineers AIChE J, 2016« less

Taipower`s radioactive waste management program

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

Lee, B.C.C.

1996-09-01

Nuclear safety and radioactive waste management are the two major concerns of nuclear power in Taiwan. Recognizing that it is an issue imbued with political and social-economic concerns, Taipower has established an integrated nuclear backend management system and its associated financial and mechanism. For LLW, the Orchid Island storage facility will play an important role in bridging the gap between on-site storage and final disposal of LLW. Also, on-site interim storage of spent fuel for 40 years or longer will provide Taipower with ample time and flexibility to adopt the suitable alternative of direct disposal or reprocessing. In other words,more » by so exercising interim storage option, Taipower will be in a comfortable position to safely and permanently dispose of radwaste without unduly forgoing the opportunities of adopting better technologies or alternatives. Furthermore, Taipower will spare no efforts to communicate with the general public and make her nuclear backend management activities accountable to them.« less

Electrochemical Impedance Spectroscopic Study on Eu 2+ and Sr 2+ Using Liquid Metal Cathodes in Molten Chlorides

NASA Astrophysics Data System (ADS)

Matsumiya, Masahiko; Takagi, Ryuzo

2000-08-01

For the pyrochemical reprocessing of spent metallic nuclear fuels in molten salt baths it is important to investigate the behavior of the electrochemically negative elements Eu and Sr, which are significant fission products. Voltammetric and chronopotentiometric studies have shown that the reduction of Eu 2+ and Sr 2+ on liquid Pb cathodes in molten chloride baths at 1073 K follows the alloy formation reaction: Eu 2+ + 2e- + 3Pb → EuPb 3 and Sr 2+ + 2e- + 3Pb → SrPb 3 . In the present work these alloy formation reactions were studiedby electrochemical impedance spectroscopy. Analysis of the spectra showed that the electronic exchange of Eu 2+ /Eu and Sr 2+ /Sr is quasi-re-versible. Moreover, the experimental results allowed the determination of the kinetic parameters of EU 2+ /EU and Sr 2+ /Sr, the diffusion coefficients of these species in molten chloride baths, and also the diffusion layer thickness.

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

Rack for storing spent nuclear fuel elements

DOEpatents

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

1978-06-20

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-17

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

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

Code of Federal Regulations, 2010 CFR

2010-01-01

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

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

Code of Federal Regulations, 2011 CFR

2011-01-01

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

75 FR 53353 - Notice of Availability of Final Interim Staff Guidance Document No. 25 “Pressure and Helium...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-31

... Guidance Document No. 25 ``Pressure and Helium Leakage Testing of the Confinement Boundary of Spent Fuel...: The Division of Spent Fuel Storage and Transportation (SFST) of the Office of Nuclear Materials Safety... Helium Leakage Testing of the Confinement Boundary of Spent Fuel Dry Storage Systems.'' This ISG...

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

Code of Federal Regulations, 2011 CFR

2011-01-01

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

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

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

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

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

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)

Hydrogen generation having CO₂ removal with steam reforming

DOEpatents

Kandaswamy, Duraiswamy; Chellappa, Anand S.; Knobbe, Mack

2015-07-28

A method for producing hydrogen using fuel cell off gases, the method feeding hydrocarbon fuel to a sulfur adsorbent to produce a desulfurized fuel and a spent sulfur adsorbent; feeding said desulfurized fuel and water to an adsorption enhanced reformer that comprises of a plurality of reforming chambers or compartments; reforming said desulfurized fuel in the presence of a one or more of a reforming catalyst and one or more of a CO₂ adsorbent to produce hydrogen and a spent CO₂ adsorbent; feeding said hydrogen to the anode side of the fuel cell; regenerating said spent CO₂ adsorbents using the fuel cell cathode off-gases, producing a flow of hydrogen by cycling between said plurality of reforming chambers or compartments in a predetermined timing sequence; and, replacing the spent sulfur adsorbent with a fresh sulfur adsorbent at a predetermined time.

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

DOEpatents

Reichner, P.; Dollard, W.J.

1991-01-08

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

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

Optimization of spent fuel pool weir gate driving mechanism

NASA Astrophysics Data System (ADS)

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

2018-04-01

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

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

Sarta, Jose A.; Castiblanco, Luis A

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-28

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

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

Code of Federal Regulations, 2011 CFR

2011-01-01

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

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

Code of Federal Regulations, 2010 CFR

2010-01-01

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

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

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

Not Available

1992-07-01

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

Assess How Changes in Fuel Cycle Operation Impact Safeguards

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

Tobin, Stephen Joseph; Adigun, Babatunde John; Fugate, Michael Lynn

Since the beginning of commercial nuclear power generation in the 1960s, the ability of researchers to understand and control the isotopic content of spent fuel has improved. It is therefore not surprising that both fuel assembly design and fuel assembly irradiation optimization have improved over the past 50+ years. It is anticipated that the burnup and isotopics of the spent fuel should exhibit less variation over the decades as reactor operators irradiate each assembly to the optimum amount. In contrast, older spent fuel is anticipated to vary more in burnup and resulting isotopics for a given initial enrichment. Modern fuelmore » therefore should be more uniform in composition, and thus, measured safeguards results should be easier to interpret than results from older spent fuel. With spent fuel ponds filling up, interim and long-­term storage of spent fuel will need to be addressed. Additionally after long periods of storage, spent fuel is no longer self-­protecting and, as such, the IAEA will categorize it as more attractive; in approximately 20 years many of the assemblies from early commercial cores will no longer be considered self-­protecting. This study will assess how more recent changes in the reactor operation could impact the interpretation of safeguards measurements. The status quo for spent fuel assay in the safeguards context is that the overwhelming majority of spent fuel assemblies are not measured in a quantitative way except for those assemblies about to be loaded into a difficult or impossible to access location (dry storage or, in the future, a repository). In other words, when the assembly is still accessible to a state actor, or an insider, when it is cooling in a pool, the inspectorate does not have a measurement database that could assist them in re-­verifying the integrity of that assembly. The spent fuel safeguards regime would be strengthened if spent fuel assemblies were measured from discharge to loading into a difficult or impossible to access location. The primary driver for suggesting this shift in approach is the change in robotic technology and information technology in general. It should be possible, with minimal impact to the facility, to measure each assembly every time that it is moved in the pool, with the first measurements being made at discharge. The following conclusions were reached: The total neutron count rate can be accurately predicted at any future moment in time based upon the measured count rate at discharge, provided the initial enrichment and burnup of the assembly is known at discharge. It is expected that the total neutron count rate measured at discharge will be indicative of the initial enrichment and burnup of that assembly. If the automated robot were to focus on measuring the assemblies in the rack without moving them, the time available would increase immensely.« less

SPECTROSCOPIC ONLINE MONITORING FOR PROCESS CONTROL AND SAFEGUARDING OF RADIOCHEMICAL STREAMS

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

Bryan, Samuel A.; Levitskaia, Tatiana G.

2013-09-29

There is a renewed interest worldwide to promote the use of nuclear power and close the nuclear fuel cycle. The long term successful use of nuclear power is critically dependent upon adequate and safe processing and disposition of the used nuclear fuel. Liquid-liquid extraction is a separation technique commonly employed for the processing of the dissolved used nuclear fuel. 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. This paper summarizes application of the absorption and vibrational spectroscopicmore » techniques supplemented by physicochemical measurements for radiochemical process monitoring. In this context, our team experimentally assessed the potential of Raman and spectrophotometric techniques for online real-time monitoring of the U(VI)/nitrate ion/nitric acid and Pu(IV)/Np(V)/Nd(III), respectively, in solutions relevant to spent fuel reprocessing. These techniques demonstrate robust performance in the repetitive batch measurements of each analyte in a wide concentration range using simulant and commercial dissolved spent fuel solutions. Spectroscopic measurements served as training sets for the multivariate data analysis to obtain partial least squares predictive models, which were validated using on-line centrifugal contactor extraction tests. Satisfactory prediction of the analytes concentrations in these preliminary experiments warrants further development of the spectroscopy-based methods for radiochemical process control and safeguarding. Additionally, the ability to identify material intentionally diverted from a liquid-liquid extraction contactor system was successfully tested using on-line process monitoring as a means to detect the amount of material diverted. A chemical diversion and detection from a liquid-liquid extraction scheme was demonstrated using a centrifugal contactor system operating with the simulant PUREX extraction system of Nd(NO3)3/nitric acid aqueous phase and TBP/n-dodecane organic phase. During a continuous extraction experiment, a portion of the feed from a counter-current extraction system was diverted while the spectroscopic on-line process monitoring system was simultaneously measuring the feed, raffinate and organic products streams. The amount observed to be diverted by on-line spectroscopic process monitoring was in excellent agreement with values based from the known mass of sample directly taken (diverted) from system feed solution.« less

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

DOE PAGES

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

2017-03-27

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

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

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

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

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

Characterization of Actinides Complexed to Nuclear Fuel Constituents Using ESI-MS.

PubMed

McDonald, Luther W; Campbell, James A; Vercouter, Thomas; Clark, Sue B

2016-03-01

Electrospray ionization-mass spectrometry (ESI-MS) was tested for its use in monitoring spent nuclear fuel (SNF) constituents including U, Pu, dibutyl phosphate (DBP), and tributyl phosphate (TBP). Both positive and negative ion modes were used to evaluate the speciation of U and Pu with TBP and DBP. Furthermore, apparent stability constants were determined for U complexed to TBP and DBP. In positive ion mode, TBP produced a strong signal with and without complexation to U or Pu, but, in negative ion mode, no TBP, U-TBP, or Pu-TBP complexes were observed. Apparent stability constants were determined for [UO2(NO3)2(TBP)2], [UO2(NO3)2(H2O)(TBP)2], and [UO2(NO3)2(TBP)3]. In contrast DBP, U-DBP, and Pu-DBP complexes were observed in both positive and negative ion modes. Apparent stability constants were determined for the species [UO2(DBP)], [UO2(DBP)3], and [UO2(DBP)4]. Analyzing mixtures of U or Pu with TBP and DBP yielded the formation of ternary complexes whose stoichiometry was directly related to the ratio of TBP to DBP. The ESI-MS protocols used in this study will further demonstrate the utility of ESI-MS and its applicability to process control monitoring in SNF reprocessing facilities.

Final environmental impact statement. Management of commercially generated radioactive waste. Volume 3. Public comments hearing board report

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

Not Available

1980-10-01

This EIS analyzes the significant environmental impacts that could occur if various technologies for management and disposal of high-level and transuranic wastes from commercial nuclear power reactors were to be developed and implemented. This EIS will serve as the environmental input for the decision on which technology, or technologies, will be emphasized in further research and development activities in the commercial waste management program. The action proposed in this EIS is to (1) adopt a national strategy to develop mined geologic repositories for disposal of commercially generated high-level and transuranic radioactive waste (while continuing to examine subseabed and very deepmore » hole disposal as potential backup technologies) and (2) conduct a R and D program to develop such facilities and the necessary technology to ensure the safe long-term containment and isolation of these wastes. The Department has considered in this statement: development of conventionally mined deep geologic repositories for disposal of spent fuel from nuclear power reactors and/or radioactive fuel reprocessing wastes; balanced development of several alternative disposal methods; and no waste disposal action. This volume contains written public comments and hearing board responses and reports offered on the draft statement.« less

Cosmic ray muons for spent nuclear fuel monitoring

NASA Astrophysics Data System (ADS)

Chatzidakis, Stylianos

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

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

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

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

Kalinina, Elena; Busch, Ingrid Karin

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

Uranium oxide fuel cycle analysis in VVER-1000 with VISTA simulation code

NASA Astrophysics Data System (ADS)

Mirekhtiary, Seyedeh Fatemeh; Abbasi, Akbar

2018-02-01

The VVER-1000 Nuclear power plant generates about 20-25 tons of spent fuel per year. In this research, the fuel transmutation of Uranium Oxide (UOX) fuel was calculated by using of nuclear fuel cycle simulation system (VISTA) code. In this simulation, we evaluated the back end components fuel cycle. The back end component calculations are Spent Fuel (SF), Actinide Inventory (AI) and Fission Product (FP) radioisotopes. The SF, AI and FP values were obtained 23.792178 ton/y, 22.811139 ton/y, 0.981039 ton/y, respectively. The obtained value of spent fuel, major actinide, and minor actinide and fission products were 23.8 ton/year, 22.795 ton/year, 0.024 ton/year and 0.981 ton/year, respectively.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-06

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-02

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

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...

Shipment of spent nuclear fuel from U.S. Navy ships and submarines to the Idaho National Engineering Laboratory (INEL). Hearing before the Subcommittee on Nuclear Deterrence, Arms Control and Defense Intelligence of the Committee on Armed Services, United States Senate, One Hundred Third Congress, First Session, July 28, 1993

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

NONE

1994-12-31

The purpose of the hearing was to review the impact of the U.S. District Court of Idaho ruling prohibiting receipt of spent nuclear fuel by the Department of Energy (DOE). The court`s ruling enjoined the DOE from receiving spent nuclear fuel, including nuclear fuel from naval surface ships and submarines, at the Idaho National Engineering Laboratory until such time as the DOE completes an environmental impact statement on the transportation, shipment, processing, and storage of spent fuel. Statements of government officials are included. The text of the Court ruling is also included.

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

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

Trond Bjornard; Philip C. Durst

2012-05-01

This document summarizes the requirements and best practices for implementing international nuclear safeguards at independent spent fuel storage installations (ISFSIs), also known as Away-from- Reactor (AFR) storage facilities. These installations may provide wet or dry storage of spent fuel, although the safeguards guidance herein focuses on dry storage facilities. In principle, the safeguards guidance applies to both wet and dry storage. The reason for focusing on dry independent spent fuel storage installations is that this is one of the fastest growing nuclear installations worldwide. Independent spent fuel storage installations are typically outside of the safeguards nuclear material balance area (MBA)more » of the reactor. They may be located on the reactor site, but are generally considered by the International Atomic Energy Agency (IAEA) and the State Regulator/SSAC to be a separate facility. The need for this guidance is becoming increasingly urgent as more and more nuclear power plants move their spent fuel from resident spent fuel ponds to independent spent fuel storage installations. The safeguards requirements and best practices described herein are also relevant to the design and construction of regional independent spent fuel storage installations that nuclear power plant operators are starting to consider in the absence of a national long-term geological spent fuel repository. The following document has been prepared in support of two of the three foundational pillars for implementing Safeguards-by-Design (SBD). These are: i) defining the relevant safeguards requirements, and ii) defining the best practices for meeting the requirements. This document was prepared with the design of the latest independent dry spent fuel storage installations in mind and was prepared specifically as an aid for designers of commercial nuclear facilities to help them understand the relevant international requirements that follow from a country’s safeguards agreement with the IAEA. If these requirements are understood at the earliest stages of facility design, it will help eliminate the costly retrofit of facilities that has occurred in the past to accommodate nuclear safeguards, and will help the IAEA implement nuclear safeguards worldwide, especially in countries building their first nuclear facilities. It is also hoped that this guidance document will promote discussion between the IAEA, State Regulator/SSAC, Project Design Team, and Facility Owner/Operator at an early stage to ensure that new ISFSIs will be effectively and efficiently safeguarded. This is intended to be a living document, since the international nuclear safeguards requirements may be subject to revision over time. More importantly, the practices by which the requirements are met are continuously modernized by the IAEA and facility operators for greater efficiency and cost effectiveness. As these improvements are made, it is recommended that the subject guidance document be updated and revised accordingly.« less

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

NASA Astrophysics Data System (ADS)

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

2017-05-01

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

Applications in Nuclear Energy Security

NASA Astrophysics Data System (ADS)

Sheffield, Richard

2009-05-01

A key roadblock to development of additional nuclear power capacity is a concern over management of nuclear waste. Nuclear waste is predominantly comprised of used fuel discharged from operating nuclear reactors. The roughly 100 operating US reactors currently produce about 20% of the US electricity and will create about 87,000 tons of such discharged or ``spent'' fuel over the course of their lifetimes. The long-term radioactivity of the spent fuel drives the need for deep geologic storage that remains stable for millions of years. Nearly all issues related to risks to future generations arising from long-term disposal of such spent nuclear fuel is attributable to approximately the 1% made up primarily of minor actinides. If we can reduce or eliminate this 1% of the spent fuel, then within a few hundred years the toxic nature of the spent fuel drops below that of the natural uranium ore that was originally mined for nuclear fuel. The minor actinides can be efficiently eliminated through nuclear transmutation using as a driver fast-neutrons produced by a spallation process initiated with a high-energy proton beam. This presentation will cover the system design considerations and issues of an accelerator driven transmutation system.

Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement

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

N /A

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

Waste Estimates for a Future Recycling Plant in the US Based Upon AREVA Operating Experience - 13206

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

Foare, Genevieve; Meze, Florian; Bader, Sven

2013-07-01

Estimates of process and secondary wastes produced by a recycling plant built in the U.S., which is composed of a used nuclear fuel (UNF) reprocessing facility and a mixed oxide (MOX) fuel fabrication facility, are performed as part of a U.S. Department of Energy (DOE) sponsored study [1]. In this study, a set of common inputs, assumptions, and constraints were identified to allow for comparison of these wastes between different industrial teams. AREVA produced a model of a reprocessing facility, an associated fuel fabrication facility, and waste treatment facilities to develop the results for this study. These facilities were dividedmore » into a number of discrete functional areas for which inlet and outlet flow streams were clearly identified to allow for an accurate determination of the radionuclide balance throughout the facility and the waste streams. AREVA relied primarily on its decades of experience and feedback from its La Hague (reprocessing) and MELOX (MOX fuel fabrication) commercial operating facilities in France to support this assessment. However, to perform these estimates for a U.S. facility with different regulatory requirements and to take advantage of some technological advancements, such as in the potential treatment of off-gases, some deviations from this experience were necessary. A summary of AREVA's approach and results for the recycling of 800 metric tonnes of initial heavy metal (MTIHM) of LWR UNF per year into MOX fuel under the assumptions and constraints identified for this DOE study are presented. (authors)« less

Detecting pin diversion from pressurized water reactors spent fuel assemblies

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

Ham, Young S.; Sitaraman, Shivakumar

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

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...

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

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

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

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

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

DOE PAGES

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

2016-02-26

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

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

PubMed

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

2005-01-01

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

Development of self-interrogation neutron resonance densitometry (sinrd) to measure the fissile content in nuclear fuel

NASA Astrophysics Data System (ADS)

LaFleur, Adrienne Marie

The development of non-destructive assay (NDA) capabilities to directly measure the fissile content in spent fuel is needed to improve the timely detection of the diversion of significant quantities of fissile material. Currently, the International Atomic Energy Agency (IAEA) does not have effective NDA methods to verify spent fuel and recover continuity of knowledge in the event of a containment and surveillance systems failure. This issue has become increasingly critical with the worldwide expansion of nuclear power, adoption of enhanced safeguards criteria for spent fuel verification, and recent efforts by the IAEA to incorporate an integrated safeguards regime. In order to address these issues, the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) has been developed to improve existing nuclear safeguards and material accountability measurements. The following characteristics of SINRD were analyzed: (1) ability to measure the fissile content in Light Water Reactors (LWR) fuel assemblies and (2) sensitivity and penetrability of SINRD to the removal of fuel pins from an assembly. The Monte Carlo Neutral Particle eXtended (MCNPX) transport code was used to simulate SINRD for different geometries. Experimental measurements were also performed with SINRD and were compared to MCNPX simulations of the experiment to verify the accuracy of the MCNPX model of SINRD. Based on the results from these simulations and measurements, we have concluded that SINRD provides a number of improvements over current IAEA verification methods. These improvements include: (1) SINRD provides absolute measurements of burnup independent of the operator's declaration. (2) SINRD is sensitive to pin removal over the entire burnup range and can verify the diversion of 6% of fuel pins within 3o from LWR spent LEU and MOX fuel. (3) SINRD is insensitive to the boron concentration and initial fuel enrichment and can therefore be used at multiple spent fuel storage facilities. (4) The calibration of SINRD at one reactor facility carries over to reactor sites in different countries because it uses the ratio of fission chambers (FCs) that are not facility dependent. (5) SINRD can distinguish fresh and 1-cycle spent MOX fuel from 3- and 4-cycles spent LEU fuel without using reactor burnup codes.

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

DOEpatents

Reichner, Philip; Dollard, Walter J.

1991-01-01

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

Spent Nuclear Fuel Transport Reliability Study

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

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

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

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

NASA Astrophysics Data System (ADS)

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

2005-02-01

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

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...

Transportation of spent MTR fuels

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

Raisonnier, D.

1997-08-01

This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs.

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

Spent Nuclear Fuel Alternative Technology Decision Analysis

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

Shedrow, C.B.

1999-11-29

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

Continuous process electrorefiner

DOEpatents

Herceg, Joseph E [Naperville, IL; Saiveau, James G [Hickory Hills, IL; Krajtl, Lubomir [Woodridge, IL

2006-08-29

A new device is provided for the electrorefining of uranium in spent metallic nuclear fuels by the separation of unreacted zirconium, noble metal fission products, transuranic elements, and uranium from spent fuel rods. The process comprises an electrorefiner cell. The cell includes a drum-shaped cathode horizontally immersed about half-way into an electrolyte salt bath. A conveyor belt comprising segmented perforated metal plates transports spent fuel into the salt bath. The anode comprises the conveyor belt, the containment vessel, and the spent fuel. Uranium and transuranic elements such as plutonium (Pu) are oxidized at the anode, and, subsequently, the uranium is reduced to uranium metal at the cathode. A mechanical cutter above the surface of the salt bath removes the deposited uranium metal from the cathode.

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

NASA Astrophysics Data System (ADS)

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

2017-11-01

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

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.

Advancing the Fork detector for quantitative spent nuclear fuel verification

DOE PAGES

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

2018-01-31

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

Advancing the Fork detector for quantitative spent nuclear fuel verification

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

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

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

Advancing the Fork detector for quantitative spent nuclear fuel verification

NASA Astrophysics Data System (ADS)

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

2018-04-01

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

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

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

Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy

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

Properties of the LiCl-KCl-Li2O system as operating medium for pyro-chemical reprocessing of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Mullabaev, Albert; Tkacheva, Olga; Shishkin, Vladimir; Kovrov, Vadim; Zaikov, Yuriy; Sukhanov, Leonid; Mochalov, Yuriy

2018-03-01

Crystallization temperatures (liquidus and solidus) in the LiCl-Li2O and (LiCl-KCl)-Li2O systems with the KCl content of 10 and 20 mol.% were obtained with independent methods of thermal analysis using cooling curves, isothermal saturation, and differential scanning calorimetry. The linear sweep voltammetry was applied to control the time of the equilibrium establishment in the molten system after the Li2O addition, which depended on the composition of the base melt and the concentration of Li2O. The fragments of the binary LiCl-Li2O and quazi-binary [LiCl-KCl(10 mol.%)]-Li2O and [LiCl-KCl(20 mol.%)]-Li2O phase diagrams in the Li2O concentration range from 0 to 12 mol.% were obtained. The KCl presence in the LiCl-KCl-Li2O molten mixture in the amount of 10 and 20 mol.% reduces the liquidus temperature by 30 and 80°, respectively, but the region of the homogeneous molten state of the system is considerably narrowed, which complicates its practical application. The Li2O solubility in the molten LiCl, LiCl-KCl(10 mol.%) and LiCl-KCl(20 mol.%) decreases with increasing the KCl content and is equal to 11.5, 7.7 and 3.9 mol.% at 650°С, respectively. The LiCl-KCl melt with 10 mol.% KCl can be recommended for practical use as a medium for the SNF pyro-chemical reprocessing at temperature below 700 °C.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-17

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-05

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

Used fuel extended storage security and safeguards by design roadmap

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

Durbin, Samuel G.; Lindgren, Eric Richard; Jones, Robert

2016-05-01

In the United States, spent nuclear fuel (SNF) is safely and securely stored in spent fuel pools and dry storage casks. The available capacity in spent fuel pools across the nuclear fleet has nearly reached a steady state value. The excess SNF continues to be loaded in dry storage casks. Fuel is expected to remain in dry storage for periods beyond the initial dry cask certification period of 20 years. Recent licensing renewals have approved an additional 40 years. This report identifies the current requirements and evaluation techniques associated with the safeguards and security of SNF dry cask storage. Amore » set of knowledge gaps is identified in the current approaches. Finally, this roadmap identifies known knowledge gaps and provides a research path to deliver the tools and models needed to close the gaps and allow the optimization of the security and safeguards approaches for an interim spent fuel facility over the lifetime of the storage site.« less

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

DOE PAGES

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

2016-10-22

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

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

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

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

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

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

NASA Astrophysics Data System (ADS)

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

2017-01-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-02

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

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

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

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

Christian, Jerry D.; Sterbentz, James W.

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

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

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

Kelly Cummins; Igor Bolshinsky; Ken Allen

2009-07-01

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

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

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

None

1978-01-31

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

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

DOEpatents

Ackerman, John P.; Miller, William E.

1989-01-01

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

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

DOEpatents

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

1987-11-05

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

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

NASA Astrophysics Data System (ADS)

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

2013-11-01

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

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.

10 CFR 110.40 - Commission review.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Canada. (5) An export involving assistance to end uses related to isotope separation, chemical reprocessing, heavy water production, advanced reactors, or the fabrication of nuclear fuel containing...

10 CFR 961.1 - Purpose.

Code of Federal Regulations, 2010 CFR

2010-01-01

... STANDARD CONTRACT FOR DISPOSAL OF SPENT NUCLEAR FUEL AND/OR HIGH-LEVEL RADIOACTIVE WASTE General § 961.1... fuel (SNF) and high-level radioactive waste (HLW) as provided in section 302 of the Nuclear Waste... title to, transport, and dispose of spent nuclear fuel and/or high-level radioactive waste delivered to...

10 CFR 72.120 - General considerations.

Code of Federal Regulations, 2010 CFR

2010-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design... reactor-related GTCC waste in an ISFSI or to store spent fuel, high-level radioactive waste, or reactor... be designed to store spent fuel and/or solid reactor-related GTCC waste. (1) Reactor-related GTCC...

Nuclear Power Plant Security and Vulnerabilities

DTIC Science & Technology

2009-03-18

Commercial Spent Nuclear Fuel Storage , Public Report...systems that prevent hot nuclear fuel from melting even after the chain reaction has stopped, and storage facilities for highly radioactive spent nuclear ... nuclear fuel cycle facilities must defend against to prevent radiological sabotage and theft of strategic special nuclear material. NRC licensees use

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.

78 FR 77736 - Sunshine Act Meetings

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-24

... a.m. Briefing on Spent Fuel Pool Safety and Consideration of Expedited Transfer of Spent Fuel to Dry... Disposition Fukushima Near-Term Task Force (NTTF) Recommendation 1 on Improving NRC's Regulatory Framework...

Global Uranium And Thorium Resources: Are They Adequate To Satisfy Demand Over The Next Half Century?

NASA Astrophysics Data System (ADS)

Lambert, I. B.

2012-04-01

This presentation will consider the adequacy of global uranium and thorium resources to meet realistic nuclear power demand scenarios over the next half century. It is presented on behalf of, and based on evaluations by, the Uranium Group - a joint initiative of the OECD Nuclear Energy Agency and the International Atomic Energy Agency, of which the author is a Vice Chair. The Uranium Group produces a biennial report on Uranium Resources, Production and Demand based on information from some 40 countries involved in the nuclear fuel cycle, which also briefly reviews thorium resources. Uranium: In 2008, world production of uranium amounted to almost 44,000 tonnes (tU). This supplied approximately three-quarters of world reactor requirements (approx. 59,000 tU), the remainder being met by previously mined uranium (so-called secondary sources). Information on availability of secondary sources - which include uranium from excess inventories, dismantling nuclear warheads, tails and spent fuel reprocessing - is incomplete, but such sources are expected to decrease in market importance after 2013. In 2008, the total world Reasonably Assured plus Inferred Resources of uranium (recoverable at less than 130/kgU) amounted to 5.4 million tonnes. In addition, it is clear that there are vast amounts of uranium recoverable at higher costs in known deposits, plus many as yet undiscovered deposits. The Uranium Group has concluded that the uranium resource base is more than adequate to meet projected high-case requirements for nuclear power for at least half a century. This conclusion does not assume increasing replacement of uranium by fuels from reprocessing current reactor wastes, or by thorium, nor greater reactor efficiencies, which are likely to ameliorate future uranium demand. However, progressively increasing quantities of uranium will need to be mined, against a backdrop of the relatively small number of producing facilities around the world, geopolitical uncertainties and strong opposition to growth of nuclear power in a number of quarters - it is vital that the market provides incentives for exploration and development of environmentally sustainable mining operations. Thorium: World Reasonably Assured plus Inferred Resources of thorium are estimated at over 2.2 million tonnes, in hard rock and heavy mineral sand deposits. At least double this amount is considered to occur in as yet undiscovered thorium deposits. Currently, demand for thorium is insignificant, but even a major shift to thorium-fueled reactors would not make significant inroads into the huge resource base over the next half century.

Spent fuel container alignment device and method

DOEpatents

Jones, Stewart D.; Chapek, George V.

1996-01-01

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

Active Interrogation for Spent Fuel

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

Swinhoe, Martyn Thomas; Dougan, Arden

2015-11-05

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

78 FR 76331 - Sunshine Act Meeting Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-17

... a.m. Briefing on Spent Fuel Pool Safety and Consideration of Expedited Transfer of Spent Fuel to Dry... Recommendations to Disposition Fukushima Near-Term Task Force (NTTF) Recommendation 1 on Improving NRC's...

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

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

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

1995-12-31

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

Determination of activity coefficient of lanthanum chloride in molten LiCl-KCl eutectic salt as a function of cesium chloride and lanthanum chloride concentrations using electromotive force measurements

NASA Astrophysics Data System (ADS)

Bagri, Prashant; Simpson, Michael F.

2016-12-01

The thermodynamic behavior of lanthanides in molten salt systems is of significant scientific interest for the spent fuel reprocessing of Generation IV reactors. In this study, the apparent standard reduction potential (apparent potential) and activity coefficient of LaCl3 were determined in a molten salt solution of eutectic LiCl-KCl as a function of concentration of LaCl3. The effect of adding up to 1.40 mol % CsCl was also investigated. These properties were determined by measuring the open circuit potential of the La-La(III) redox couple in a high temperature molten salt electrochemical cell. Both the apparent potential and activity coefficient exhibited a strong dependence on concentration. A low concentration (0.69 mol %) of CsCl had no significant effect on the measured properties, while a higher concentration (1.40 mol %) of CsCl caused an increase (become more positive) in the apparent potential and activity coefficient at the higher range of LaCl3 concentrations.

Radiocarbon behaviour in seawater and the brown algae Fucus serratus in the vicinity of the COGEMA La Hague spent fuel reprocessing plant (Goury)--France.

PubMed

Douville, Eric; Fiévet, Bruno; Germain, Pierre; Fournier, Marc

2004-01-01

Extensive studies of the radiocarbon (14C) distribution and transfer in the marine environment of the North-Cotentin peninsula and along the English Channel have been carried out. The main aims of these studies have been to estimate the spatial and temporal variation of the 14C concentration in seawater and to calculate 14C concentration factors for some biological species. Such information will be helpful in order to calculate precisely radiation doses to humans. First results obtained in the vicinity of the COGEMA La Hague nuclear plant (Goury) indicate a 14C labelling of the dissolved inorganic carbon (DIC) in seawater (8.0-26.2 Bq.m(-3)) and a tight relationship between the 14C in the liquid releases from the plant and the 14C concentrations in DIC. The particulate organic carbon (POC) is also labelled. The concentration factor calculations for the brown algae (Fucus serratus) sampled from Goury, and also along the English Channel, give 14C values around 3000 Bq.kg(-1) fresh weight / Bq.L(-1).

Experimental investigation of differential confinement effects in a rotating helicon plasma

NASA Astrophysics Data System (ADS)

Gueroult, Renaud; Evans, Eugene; Zweben, Stewart J.; Fisch, Nathaniel J.; Levinton, Fred

2014-10-01

Although plasmas have long been considered for isotope separation, challenges presented by nuclear waste remediation and nuclear spent fuel reprocessing have recently sparked a renewed interest for high-throughput plasma based mass separation techniques. Different filter concepts relying on rotating plasmas have been proposed to address these needs. However, one of the challenges common to these concepts is the need to control the plasma rotation profile, which is generally assumed to be provided by means of dedicated electrodes. An experimental effort aiming to evaluate the practicality of these plasma filter concepts has recently been started at PPPL. For this purpose, a linear helicon plasma source is used in combination with concentric ring electrodes. Preliminary biasing experiments results indicate floating potential profiles locally suitable for mass discrimination for different gas mixtures (Ar/Ne, Ar/N2, Ar/Kr). Radially resolved spectroscopic measurements and neutral gas composition analysis at two different axial positions are being planned to assess the mass separation effect. Work supported by US DOE under Contract No. DE-AC02-09CH11466.

Adsorption of iodine on hydrogen-reduced silver-exchanged mordenite: Experiments and modeling

DOE PAGES

Nan, Yue; Tavlarides, Lawrence L.; DePaoli, David W.

2016-08-03

The adsorption process of iodine, a major volatile radionuclide in the off-gas streams of spent nuclear fuel reprocessing, on hydrogen-reduced silver-exchanged mordenite (Ag 0Z) was studied at the micro-scale. The gas-solid mass transfer and reaction involved in the adsorption process were investigated and evaluated with appropriate models. Optimal conditions for reducing the silver-exchanged mordenite (AgZ) in a hydrogen stream were determined. Kinetic and equilibrium data of iodine adsorption on Ag 0Z were obtained by performing single-layer adsorption experiments with experimental systems of high precision at 373–473 K over various iodine concentrations. Results indicate approximately 91% to 97% of the iodinemore » adsorption was through the silver-iodine reaction. The effect of temperature on the iodine loading capacity of Ag 0Z was discussed. In conclusion, the Shrinking Core model describes the data well, and the primary rate controlling mechanisms were macro-pore diffusion and silver-iodine reaction. © 2016 American Institute of Chemical Engineers AIChE J, 2016« less

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...

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

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

Ludwick, J D; Moore, E B

1984-01-01

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

Performance testing and analyses of the VSC-17 ventilated concrete cask. Final report

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

McKinnon, M.A.; Dodge, R.E.; Schmitt, R.C.

1992-05-01

This document details performance test which was conducted on a Pacific Sierra Nuclear VSC-17 ventilated concrete storage cask configured for pressurized-water reactor (PWR) spent fuel. The performance test consisted of loading the VSC-17 cask with 17 canisters of consolidated PWR spent fuel from Virginia Power`s Surry and Florida Power & Light Turkey Point reactors. Cask surface, concrete, air channel surfaces, and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in a vertical cask orientation. Data on spent fuel integrity were also obtained.

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

NASA Astrophysics Data System (ADS)

Acır, Adem; Altunok, Taner

2010-10-01

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

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

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

Horne, Michael; Clough, Malcolm

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

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

Characterization of Actinides Complexed to Nuclear Fuel Constituents Using ESI-MS

DOE PAGES

McDonald, Luther W.; Campbell, James A.; Vercouter, Thomas; ...

2016-03-01

Electrospray ionization-mass spectrometry (ESI-MS) was tested for its use in monitoring spent nuclear fuel (SNF) constituents including U, Pu, dibutyl phosphate (DBP), and tributyl phosphate (TBP). Both positive and negative ion modes were used to evaluate the speciation of U and Pu with TBP and DBP. Furthermore, apparent stability constants were determined for U complexed to TBP and DBP. In positive ion mode, TBP produced a strong signal with and without complexation to U or Pu, but, in negative ion mode, no TBP, U-TBP, or Pu-TBP complexes were observed. Apparent stability constants were determined for [UO 2(NO 3) 2(TBP) 2],more » [UO 2(NO 3) 2(H 2O)(TBP) 2], and [UO 2(NO 3) 2(TBP) 3]. In contrast DBP, U-DBP, and Pu-DBP complexes were observed in both positive and negative ion modes. Apparent stability constants were determined for the species [UO 2(DBP)], [UO 2(DBP) 3], and [UO 2(DBP) 4]. Analyzing mixtures of U or Pu with TBP and DBP yielded the formation of ternary complexes whose stoichiometry was directly related to the ratio of TBP to DBP. The ESI-MS protocols used in this study will further demonstrate the utility of ESI-MS and its applicability to process control monitoring in SNF reprocessing facilities.« less

YUCCA Mountain Project - Argonne National Laboratory, Annual Progress Report, FY 1997 for activity WP 1221 unsaturated drip condition testing of spent fuel and unsaturated dissolution tests of glass.

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

Bates, J. K.; Buck, E. C.; Emery, J. W.

1998-09-18

This document reports on the work done by the Nuclear Waste Management Section of the Chemical Technology Division of Argonne National Laboratory in the period of October 1996 through September 1997. Studies have been performed to evaluate the behavior of nuclear waste glass and spent fuel samples under the unsaturated conditions (low-volume water contact) that are likely to exist in the Yucca Mountain environment being considered as a potential site for a high-level waste repository. Tests with actinide-doped waste glasses, in progress for over 11 years, indicate that the transuranic element release is dominated by colloids that continuously form andmore » span from the glass surface. The nature of the colloids that form in the glass and spent fuel testing programs is being investigated by dynamic light scattering to determine the size distribution, by autoradiography to determine the chemistry, and by zeta potential to measure the electrical properties of the colloids. Tests with UO{sub 2} have been ongoing for 12 years. They show that the oxidation of UO{sub 2} occurs rapidly, and the resulting paragenetic sequence of secondary phases forming on the sample surface is similar to that observed for uranium found in natural oxidizing environments. The reaction of spent fuel samples in conditions similar to those used with UO{sub 2} have been in progress for over six years, and the results suggest that spent fuel forms many of the same alteration products as UO{sub 2}. With spent fuel, the bulk of the reaction occurs via a through-grain reaction process, although grain boundary attack is sufficient to have reacted all of the grain boundary regions in the samples. New test methods are under development to evaluate the behavior of spent fuel samples with intact cladding: the rate at which alteration and radionuclide release occurs when water penetrates fuel sections and whether the reaction causes the cladding to split. Alteration phases have been formed on fine grains of UO{sub 2} in contact with small volumes of water within a several month period when the radiolysis product H{sub 2}O{sub 2} is added to the groundwater solution. The test setup has been mocked up for operation with spent fuel in the hot-cell.« less

10 CFR 72.218 - Termination of licenses.

Code of Federal Regulations, 2011 CFR

2011-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.218 Termination of licenses. (a) The notification regarding the program for the management of spent fuel at the reactor required by § 50.54(bb) of...

10 CFR 72.218 - Termination of licenses.

Code of Federal Regulations, 2010 CFR

2010-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.218 Termination of licenses. (a) The notification regarding the program for the management of spent fuel at the reactor required by § 50.54(bb) of...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-16

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

Canister Design for Deep Borehole Disposal of Nuclear Waste

DTIC Science & Technology

2006-05-01

radioactive waste disposal (not yet released) Fortunately, transportation casks for spent fuel have already been approved, built, and used as...would allow use of the current designs for transportation casks ; or, place the fuel assemblies into the final disposal canisters 21 prior to transport ...16 Figure 1-5. Typical Spent Fuel Transportation Casks

Processing fissile material mixtures containing zirconium and/or carbon

DOEpatents

Johnson, Michael Ernest; Maloney, Martin David

2013-07-02

A method of processing spent TRIZO-coated nuclear fuel may include adding fluoride to complex zirconium present in a dissolved TRIZO-coated fuel. Complexing the zirconium with fluoride may reduce or eliminate the potential for zirconium to interfere with the extraction of uranium and/or transuranics from fission materials in the spent nuclear fuel.

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

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

None

1978-01-31

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

The North Korean nuclear dilemma.

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

Hecker, Siegfried S.

2004-01-01

The current nuclear crisis, the second one in ten years, erupted when North Korea expelled international nuclear inspectors in December 2002, then withdrew from the Nuclear Nonproliferation Treaty (NPT), and claimed to be building more nuclear weapons with the plutonium extracted from the spent fuel rods heretofore stored under international inspection. These actions were triggered by a disagreement over U.S. assertions that North Korea had violated the Agreed Framework (which froze the plutonium path to nuclear weapons to end the first crisis in 1994) by clandestinely developing uranium enrichment capabilities providing an alternative path to nuclear weapons. With Stanford Universitymore » Professor John Lewis and three other Americans, I was allowed to visit the Yongbyon Nuclear Center on Jan. 8, 2004. We toured the 5 MWe reactor, the 50 MWe reactor construction site, the spent fuel pool storage building, and the radiochemical laboratory. We concluded that North Korea has restarted its 5 MWe reactor (which produces roughly 6 kg of plutonium annually), it removed the 8000 spent fuel rods that were previously stored under IAEA safeguards from the spent fuel pool, and that it most likely extracted the 25 to 30 kg of plutonium contained in these fuel rods. Although North Korean officials showed us what they claimed was their plutonium metal product from this reprocessing campaign, we were not able to conclude definitively that it was in fact plutonium metal and that it came from the most recent reprocessing campaign. Nevertheless, our North Korean hosts demonstrated that they had the capability, the facility and requisite capacity, and the technical expertise to produce plutonium metal. On the basis of our visit, we were not able to address the issue of whether or not North Korea had a 'deterrent' as claimed - that is, we were not able to conclude that North Korea can build a nuclear device and that it can integrate nuclear devices into suitable delivery systems. However, based on the capabilities we saw, we must assume that North Korea has the capability to produce a crude nuclear device. On the matter of uranium enrichment programs, our host categorically denied that North Korea has a uranium enrichment program - he said, 'we have no program, no equipment, and no technical expertise for uranium enrichment.' The denials were not convincing at the time and since then have proven to be quite hollow by the revelations of A.Q. Khan's nuclear black market activities. There is no easy solution to the nuclear crisis in North Korea. A military strike to eliminate the nuclear facilities was never very attractive and now has been overcome by events. The principal threat is posed by a stockpile of nuclear weapons and weapons-grade plutonium. We have no way of finding where either may be hidden. A diplomatic solution remains the only path forward, but it has proven elusive. All sides have proclaimed a nuclear weapons-free Korean Peninsula as the end goal. The U.S. Government has chosen to negotiate with North Korea by means of the six-party talks. It has very clearly outlined its position of insisting on complete, verifiable, irreversible dismantlement of all North Korean nuclear programs. North Korea has offered several versions of 're-freezing' its plutonium program while still denying a uranium enrichment program. It has insisted on simultaneous and reciprocal steps to a final solution. Regardless of which diplomatic path is chosen, the scientific challenges of eliminating the North Korean nuclear weapons programs (and its associated infrastructure) in a safe, secure, and verifiable manner are immense. The North Korean program is considerably more complex and developed than the fledgling Iraqi program of 1991 and Libyan program of 2004. It is more along the lines, but more complex than that of South Africa in the early 1990s. Actions taken or not taken by the North Koreans at their nuclear facilities during the course of the ongoing diplomatic discussions are key to whether or not the nuclear program can be eliminated safely and securely, and they will greatly influence the price tag for such operations. Moreover, they will determine whether or not one can verify complete elimination. Hence, cooperation of the North Koreans now and during the dismantlement and elimination stages is crucial. Technical discussions among specialists, perhaps within the framework of the working groups of the six-party talks, could be very productive in setting the stage for an effective, verifiable elimination of North Korea's nuclear weapons program.« less

Audio Script for Information Center Transportation Display

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

NA

2003-05-26

Can waste be transported safely to Yucca Mountain? Both the Department of Energy and the Nuclear Regulatory Commission have found that spent nuclear fuel can be shipped safely and securely. In fact, over the last 30 years there have been more than 2,700 shipments of spent nuclear fuel traveling more than 1.7 million miles, and there has never been a release of radioactive material harmful to the public or the environment--not one. Spent nuclear fuel is a solid material--it cannot leak, burn, or explode. The shipping containers, called casks, are the most robust in the transportation industry and must bemore » certified by the Nuclear Regulatory Commission. They are designed to protect public health and safety under normal and severe accident conditions. Typically, every ton of shipped spent fuel is contained within approximately 4 tons of protective shielding and structural materials. How many shipments would be made to Yucca Mountain? DOE would use mainly trains and some legal-weight trucks to move spent nuclear fuel and high-level radioactive waste to Yucca Mountain. Once the repository opens, DOE estimates and average of 130 rail shipments and 45 truck shipments per year for 24 years.« less

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

PubMed

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

2013-12-01

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

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

NASA Astrophysics Data System (ADS)

Gwynne, Peter

2016-07-01

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

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

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

Koji Shirai

2006-04-01

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

Spent fuel behavior under abnormal thermal transients during dry storage

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

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

1986-01-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-06

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

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

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

Louvat, D.; Lacoste, A.C.

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

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

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

St. John, C.M.

1977-04-01

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

Neutron radiation characteristics of the IVth generation reactor spent fuel

NASA Astrophysics Data System (ADS)

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

2018-03-01

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

Systems for the Intermodal Routing of Spent Nuclear Fuel

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

Peterson, Steven K; Liu, Cheng

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

Air Shipment of Spent Nuclear Fuel from Romania to Russia

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

Igor Bolshinsky; Ken Allen; Lucian Biro

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

Process for immobilizing plutonium into vitreous ceramic waste forms

DOEpatents

Feng, Xiangdong; Einziger, Robert E.

1997-01-01

Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

Process for immobilizing plutonium into vitreous ceramic waste forms

DOEpatents

Feng, X.; Einziger, R.E.

1997-08-12

Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

Process for immobilizing plutonium into vitreous ceramic waste forms

DOEpatents

Feng, X.; Einziger, R.E.

1997-01-28

Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

Method for storing spent nuclear fuel in repositories

DOEpatents

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

1981-01-01

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

Method for storing spent nuclear fuel in repositories

DOEpatents

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

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

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.

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

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

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

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

Spent fuel cask handling at an operating nuclear power plant

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

Pal, A.C.

1988-01-01

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

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

DOE PAGES

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

2015-12-01

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

Spent fuel treatment at ANL-West

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

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

1994-12-31

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

Method of increasing the deterrent to proliferation of nuclear fuels

DOEpatents

Rampolla, Donald S.

1982-01-01

A process of recycling protactinium-231 to enhance the utilization of radioactively hot uranium-232 in nuclear fuel for the purpose of making both fresh and spent fuel more resistant to proliferation. The uranium-232 may be obtained by the irradiation of protactinium-231 which is normally found in the spent fuel rods of a thorium base nuclear reactor. The production of protactinium-231 and uranium-232 would be made possible by the use of the thorium uranium-233 fuel cycle in power reactors.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-17

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

Separator assembly for use in spent nuclear fuel shipping cask

DOEpatents

Bucholz, James A.

1983-01-01

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

Gamma and fast neutron radiation monitoring inside spent reactor fuel assemblies

NASA Astrophysics Data System (ADS)

Lakosi, L.; Tam Nguyen, C.

2007-09-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-16

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

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)

Determination of (236)U and transuranium elements in depleted uranium ammunition by alpha-spectrometry and ICP-MS.

PubMed

Desideri, D; Meli, M A; Roselli, C; Testa, C; Boulyga, S F; Becker, J S

2002-11-01

It is well known that ammunition containing depleted uranium (DU) was used by NATO during the Balkan conflict. To evaluate the origin of DU (the enrichment of natural uranium or the reprocessing of spent nuclear fuel) it is necessary to directly detect the presence of activation products ((236)U, (239)Pu, (240)Pu, (241)Am, and (237)Np) in the ammunition. In this work the analysis of actinides by alpha-spectrometry was compared with that by inductively coupled plasma mass spectrometry (ICP-MS) after selective separation of ultratraces of transuranium elements from the uranium matrix. (242)Pu and (243)Am were added to calculate the chemical yield. Plutonium was separated from uranium by extraction chromatography, using tri- n-octylamine (TNOA), with a decontamination factor higher than 10(6); after elution plutonium was determined by ICP-MS ((239)Pu and (240)Pu) and alpha-spectrometry ((239+240)Pu) after electroplating. The concentration of Pu in two DU penetrator samples was 7 x 10(-12) g g(-1) and 2 x 10(-11) g g(-1). The (240)Pu/(239)Pu isotope ratio in one penetrator sample (0.12+/-0.04) was significantly lower than the (240)Pu/(239)Pu ratios found in two soil samples from Kosovo (0.35+/-0.10 and 0.27+/-0.07). (241)Am was separated by extraction chromatography, using di(2-ethylhexyl)phosphoric acid (HDEHP), with a decontamination factor as high as 10(7). The concentration of (241)Am in the penetrator samples was 2.7 x 10(-14) g g(-1) and <9.4 x 10(-15) g g(-1). In addition (237)Np was detected at ultratrace levels. In general, ICP-MS and alpha-spectrometry results were in good agreement. The presence of anthropogenic radionuclides ((236)U, (239)Pu,(240)Pu, (241)Am, and (237)Np) in the penetrators indicates that at least part of the uranium originated from the reprocessing of nuclear fuel. Because the concentrations of radionuclides are very low, their radiotoxicological effect is negligible.

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.

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.

78 FR 78165 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-26

... Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9 AGENCY: Nuclear Regulatory... storage regulations by revising the Holtec International HI-STORM 100 Cask System listing within the...

Array Detector Modules for Spent Fuel Verification

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

Bolotnikov, Aleksey

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

Dissolution of spent nuclear fuel in carbonate-peroxide solution

NASA Astrophysics Data System (ADS)

Soderquist, Chuck; Hanson, Brady

2010-01-01

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

75 FR 11375 - Revision of Fee Schedules; Fee Recovery for FY 2010

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-10

... Spent Fuel Storage/Reactor Decommissioning..... 2.7 0.2 0.2 Test and Research Reactors 0.2 0.0 0.0 Fuel... categories of licenses. The FY 2009 fee is also shown for comparative purposes. Table V--Rebaselined Annual...) Spent Fuel Storage/Reactor 122,000 143,000 Decommissioning Test and Research Reactors (Non-power 87,600...

75 FR 11566 - Firstenergy Nuclear Operating Company, Firstenergy Nuclear Generation Corp., Ohio Edison Company...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-11

... with zero soluble boron in the spent fuel pool, and that k eff remains less than or equal to 0.95 for the entire pool with credit for soluble boron under non-accident and accident conditions with a 95... k eff through fuel storage requirements and boron concentration controls in the spent fuel pool. The...

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

PubMed

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

2013-05-01

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

Romania: Brand-New Engineering Solutions

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

Ken Allen; Lucian Biro; Nicolae Zamfir

The HEU spent nuclear fuel transport from Romania was a pilot project in the framework of the Russian Research Reactor Fuel Return Program (RRRFR), being the first fully certified spent nuclear fuel shipment by air. The successful implementation of the Romanian shipment also brought various new technology in the program, further used by other participating countries. Until 2009, the RRRFR program repatriated to the Russian Federation HEU spent nuclear fuel of Russian origin from many countries, like Uzbekistan, Czech Republic, Latvia, Hungary, Kazakhstan and Bulgaria. The means of transport used were various; from specialized TK-5 train for the carriage ofmore » Russian TUK-19 transport casks, to platform trains for 20 ft freight ISO containers carrying Czech Skoda VPVR/M casks; from river barge on the Danube, to vessel on the Mediterranean Sea and Atlantic Ocean. Initially, in 2005, the transport plan of the HEU spent nuclear fuel from the National Institute for R&D in Nuclear Physics and Nuclear Engineering 'Horia Hulubei' in Magurele, Romania considered a similar scheme, using the specialized TK-5 train transiting Ukraine to the destination point in the Russian Federation, or, as an alternative, using the means and route of the spent nuclear fuel periodically shipped from the Bulgarian nuclear power plant Kosloduy (by barge on the Danube, and by train through Ukraine to the Russian Federation). Due to impossibility to reach an agreement in due time with the transit country, in February 2007 the US, Russian and Romanian project partners decided to adopt the air shipment of the spent nuclear fuel as prime option, eliminating the need for agreements with any transit countries. By this time the spent nuclear fuel inspections were completed, proving the compliance of the burn-up parameters with the international requirements for air shipments of radioactive materials. The short air route avoiding overflying of any other countries except the country of origin and the country of destination also contributed to the decision making in this issue. The efficient project management and cooperation between the three countries (Russia, Romania and USA) made possible, after two and a half years of preparation work, for the first fully certified spent nuclear fuel air shipment to take place on 29th of June 2009, from Romanian airport 'Henri Coanda' to the Russian airport 'Koltsovo' near Yekaterinburg. One day before that, after a record period of 3 weeks of preparation, another HEU cargo was shipped by air from Romanian Institute for Nuclear Research in Pitesti to Russia, containing fresh pellets and therefore making Romania the third HEU-free country in the RRRFR program.« less

Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

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

Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A., E-mail: Azizov-EA@nrcki.ru

2015-12-15

The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel canmore » be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.« less