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1

Spent nuclear fuel discharges from U.S. reactors 1994  

Science Conference Proceedings (OSTI)

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

NONE

1996-02-01T23:59:59.000Z

2

Spent nuclear fuel discharges from US reactors 1993  

SciTech Connect

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

Not Available

1995-02-01T23:59:59.000Z

3

EA-0912: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

2: Urgent-Relief Acceptance of Foreign Research Reactor Spent 2: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel EA-0912: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel SUMMARY This EA evaluates the environmental impacts of a proposal to accept 409 spent fuel elements from eight foreign research reactors in seven European countries. The spent fuel would be shipped across the ocean in spent fuel transportation casks from the country of origin to one or more United States eastern seaboard ports. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD April 22, 1994 EA-0912: Finding of No Significant Impact Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel April 22, 1994 EA-0912: Final Environmental Assessment Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

4

Development of Technical Nuclear Forensics for Spent Research Reactor Fuel  

E-Print Network (OSTI)

Pre-detonation technical nuclear forensics techniques for research reactor spent fuel were developed in a collaborative project with Savannah River National Lab ratory. An inverse analysis method was employed to reconstruct reactor parameters from a spent fuel sample using results from a radiochemical analysis. In the inverse analysis, a reactor physics code is used as a forward model. Verification and validation of different reactor physics codes was performed for usage in the inverse analysis. The verification and validation process consisted of two parts. The first is a variance analysis of Monte Carlo reactor physics burnup simulation results. The codes used in this work are MONTEBURNS and MCNPX/CINDER. Both utilize Monte Carlo transport calculations for reaction rate and flux results. Neither code has a variance analysis that will propagate through depletion steps, so a method to quantify and understand the variance propagation through these depletion calculations was developed. The second verification and validation process consisted of comparing reactor physics code output isotopic compositions to radiochemical analysis results. A sample from an Oak Ridge Research Reactor spent fuel assembly was acquired through a drilling process. This sample was then dissolved in nitric acid and diluted in three different quantities, creating three separate samples. A radiochemical analysis was completed and the results were compared to simulation outputs at different levels ofdetail. After establishing a forward model, an inverse analysis was developed to re-construct the burnup, initial uranium isotopic compositions, and cooling time of a research reactor spent fuel sample. A convergence acceleration technique was used that consisted of an analytical calculation to predict burnup, initial 235U, and 236U enrichments. The analytic calculation results may also be used stand alone or in a database search algorithm. In this work, a reactor physics code is used as a for- ward model with the analytic results as initial conditions in a numerical optimization algorithm. In the numerical analysis, the burnup and initial uranium isotopic com- positions are reconstructed until the iterative spent fuel characteristics converge with the measured data. Upon convergence of the sample’s burnup and initial uranium isotopic composition, the cooling time can be reconstructed. To reconstruct cooling time, the standard decay equation is inverted and solved for time. Two methods were developed. One method uses the converged burnup and initial uranium isotopic compositions along in a reactor depletion simulation. The second method uses an isotopic signature that does not decay out of its mass bin and has a simple production chain. An example would be 137Cs which decays into the stable 137Ba. Similar results are achieved with both methods, but extended shutdown time or time away from power results in over prediction of the cooling time. The over prediction of cooling time and comparison of different burnup reconstruction isotope results are indicator signatures of extended shutdown or time away from power. Due to dynamic operation in time and function, detailed power history reconstruction for research reactors is very challenging. Frequent variations in power, repeated variable shutdown time length, and experimentation history affect the spectrum an individual assembly is burned with such that full reactor parameter reconstruction is difficult. The results from this technical nuclear forensic analysis may be used with law enforcement, intelligence data, macroscopic and microscopic sample characteristics in a process called attribution to suggest or exclude possible sources of origin for a sample.

Sternat, Matthew 1982-

2012-12-01T23:59:59.000Z

5

Foreign Research Reactor Spent Nuclear Fuel Acceptance Program  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Global Threat Reduction Initiative: Global Threat Reduction Initiative: U.S. Nuclear Remove Program Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance 2007 DOE TEC Meeting Chuck Messick DOE/NNSA/SRS 2 Contents * Program Objective and Policy * Program implementation status * Shipment Information * Operational Logistics * Lessons Learned * Conclusion 3 U.S. Nuclear Remove Program Objective * To play a key role in the Global Threat Reduction Remove Program supporting permanent threat reduction by accepting program eligible material. * Works in conjunction with the Global Threat Reduction Convert Program to accept program eligible material as an incentive to core conversion providing a disposition path for HEU and LEU during the life of the Acceptance Program. 4 Reasons for the Policy

6

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

Science Conference Proceedings (OSTI)

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

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

2012-09-01T23:59:59.000Z

7

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

SciTech Connect

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.

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

2011-07-18T23:59:59.000Z

8

Spent nuclear fuel discharges from US reactors 1992  

SciTech Connect

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

Not Available

1994-05-05T23:59:59.000Z

9

Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

6 6 Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Reactor Sites December 2008 U.S. Department of Energy Office of Civilian Radioactive Waste Management Washington, D.C. Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel The picture on the cover is the Connecticut Yankee Independent Spent Fuel Storage Installation site in Haddam, Connecticut, with 43 dry storage NRC-licensed dual-purpose (storage and transport) casks. ii Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel EXECUTIVE SUMMARY The House Appropriations Committee Print that accompanied the Consolidated Appropriations Act, 2008, requests that the U.S. Department of Energy (the Department):

10

Nuclear nonproliferation: Concerns with US delays in accepting foregin research reactors` spent fuel  

SciTech Connect

One key US nonproliferation goal is to discourage use of highly enriched uranium fuel (HEU), which can be used to make nuclear bombs, in civilian nuclear programs worldwide. DOE`s Off-Site Fuels Policy for taking back spent HEU from foreign research reactors was allowed to expire due to environmental reasons. This report provides information on the effects of delays in renewing the Off-Site Fuels Policy on US nonproliferation goals and programs (specifically the reduced enrichment program), DOE`s efforts to renew the fuels policy, and the price to be charged to the operators of foreign reactors for DOE`s activities in taking back spent fuel.

1994-03-25T23:59:59.000Z

11

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

Science Conference Proceedings (OSTI)

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

Not Available

1994-04-01T23:59:59.000Z

12

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

SciTech Connect

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

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

2009-03-01T23:59:59.000Z

13

Storage of spent fuel from the nation`s nuclear reactors: Status, technology, and policy options  

SciTech Connect

Since the beginning of the commercial nuclear electric power industry, it has been recognized that spent nuclear reactor fuel must be able to be readily removed from the reactor vessel in the plant and safely stored on-site. The need for adjacent ready storage is first for safety. In the event of an emergency, or necessary maintenance that requires the removal of irradiated fuel from the reactor vessel, cooled reserve storage capacity for the full amount of fuel from the reactor core must be available. Also, the uranium fuel in the reactor eventually reaches the point where its heat generation is below the planned efficiency for steam production which drives the turbines and generators. It then must be replaced by fresh uranium fuel, with the ``spent fuel`` elements being removed to a safe and convenient storage location near the reactor vessel. The federal nuclear waste repository program, even without delays in the current schedule of disposal becoming available in 2003, will result in a large percentage of the 111 existing operable commercial reactors requiring expansion of their spent fuel storage capacity. How that need can and will be met raises issues of both technology and policy that will be reviewed in this report.

1989-10-01T23:59:59.000Z

14

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

SciTech Connect

This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO-type fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg[sup 235]U, and for fission product decay times of up to 20 years. Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time.

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

1996-12-31T23:59:59.000Z

15

Record of Decision for the Final EIS on Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

5091 5091 Friday May 17, 1996 Part IV Department of Energy Record of Decision for the Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel; Notice 25092 Federal Register / Vol. 61, No. 97 / Friday, May 17, 1996 / Notices DEPARTMENT OF ENERGY Record of Decision for the Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel AGENCY: Department of Energy. ACTION: Record of decision. SUMMARY: DOE, in consultation with the Department of State, has decided to implement a new foreign research reactor spent fuel acceptance policy as specified in the Preferred Alternative contained in the Final Environmental Impact Statement on a Proposed

16

DOE/EIS-0218-SA-3: Supplement Analysis for the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program (November 2004)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

SUPPLEMENT ANALYSIS FOR THE FOREIGN SUPPLEMENT ANALYSIS FOR THE FOREIGN RESEARCH REACTOR SPENT NUCLEAR FUEL ACCEPTANCE PROGRAM NOVEMBER 2004 DOE/EIS-0218-SA-3 U.S. Department of Energy National Nuclear Security Administration Washington, DC Final Supplement Analysis for the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program Final i TABLE OF CONTENTS Page 1. Introduction.............................................................................................................................................. 1 2. Background .............................................................................................................................................. 1 3. The Proposed Action ...............................................................................................................................

17

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

SciTech Connect

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, are 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.

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

1996-05-01T23:59:59.000Z

18

Foreign Research Reactor Spent Nuclear Fuel Acceptance Program  

National Nuclear Security Administration (NNSA)

rod system. * Note: Does not include the steam turbine generator portion of the power plant. - Sensitive nuclear technology: Any information (including information...

19

A compact breed and burn fast reactor using spent nuclear fuel blanket  

Science Conference Proceedings (OSTI)

A long-life breed-and-burn (B and B) type fast reactor has been investigated from the neutronics points of view. The B and B reactor has the capability to breed the fissile fuels and use the bred fuel in situ in the same reactor. In this work, feasibility of a compact sodium-cooled B and B fast reactor using spent nuclear fuel as blanket material has been studied. In order to derive a compact B and B fast reactor, a tight fuel lattice and relatively large fuel pin are used to achieve high fuel volume fraction. The core is initially loaded with an LEU (Low Enriched Uranium) fuel and a metallic fuel is used in the core. The Monte Carlo depletion has been performed for the core to see the long-term behavior of the B and B reactor. Several important parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, fission power, and fast neutron fluence, are analyzed through Monte Carlo reactor analysis. Evolution of the core fuel composition is also analyzed as a function of burnup. Although the long-life small B and B fast reactor is found to be feasible from the neutronics point of view, it is characterized to have several challenging technical issues including a very high fast neutron fluence of the structural materials. (authors)

Hartanto, D.; Kim, Y. [Korea Advanced Inst. of Science and Technology KAIST, 291 Daehak-ro, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

2012-07-01T23:59:59.000Z

20

Spent nuclear fuel shipping cask handling capabilities of commercial light water reactors  

Science Conference Proceedings (OSTI)

This report describes an evaluation of the cask handling capabilities of those reactors which are operating or under construction. A computerized data base that includes cask handling information was developed with information from the literature and utility-supplied data. The capability of each plant to receive and handle existing spent fuel shipping casks was then evaluated. Modal fractions were then calculated based on the results of these evaluations and the quantities of spent fuel projected to be generated by commercial nuclear power plants through 1998. The results indicated that all plants are capable of receiving and handling truck casks. Up to 118 out of 130 reactors (91%) could potentially handle the larger and heavier rail casks if the maximum capability of each facility is utilized. Design and analysis efforts and physical modifications to some plants would be needed to achieve this high rail percentage. These modifications would be needed to satisfy regulatory requirements, increase lifting capabilities, develop rail access, or improve other deficiencies. The remaining 12 reactors were determined to be capable of handling only the smaller truck casks. The percentage of plants that could receive and handle rail casks in the near-term would be reduced to 64%. The primary reason for a plant to be judged incapable of handling rail casks in the near-term was a lack of rail access. The remaining 36% of the plants would be limited to truck shipments. The modal fraction calculations indicated that up to 93% of the spent fuel accumulated by 1998 could be received at federal storage or disposal facilities via rail (based on each plant's maximum capabilities). If the near-term cask handling capabilities are considered, the rail percentage is reduced to 62%.

Daling, P.M.; Konzek, G.J.; Lezberg, A.J.; Votaw, E.F.; Collingham, M.I.

1985-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

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  

Science Conference Proceedings (OSTI)

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.

Not Available

1992-07-01T23:59:59.000Z

22

Storage of spent fuel from the nation's nuclear reactors: Status, technology, and policy options  

SciTech Connect

Since the beginning of the commercial nuclear electric power industry, it has been recognized that spent nuclear reactor fuel must be able to be readily removed from the reactor vessel in the plant and safely stored on-site. The need for adjacent ready storage is first for safety. In the event of an emergency, or necessary maintenance that requires the removal of irradiated fuel from the reactor vessel, cooled reserve storage capacity for the full amount of fuel from the reactor core must be available. Also, the uranium fuel in the reactor eventually reaches the point where its heat generation is below the planned efficiency for steam production which drives the turbines and generators. It then must be replaced by fresh uranium fuel, with the spent fuel'' elements being removed to a safe and convenient storage location near the reactor vessel. The federal nuclear waste repository program, even without delays in the current schedule of disposal becoming available in 2003, will result in a large percentage of the 111 existing operable commercial reactors requiring expansion of their spent fuel storage capacity. How that need can and will be met raises issues of both technology and policy that will be reviewed in this report.

1989-10-01T23:59:59.000Z

23

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

DOE Patents (OSTI)

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

Corletti, Michael M. (New Kensington, PA); Lau, Louis K. (Monroeville, PA); Schulz, Terry L. (Murrysville Boro, PA)

1993-01-01T23:59:59.000Z

24

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

DOE Patents (OSTI)

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

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

1993-12-14T23:59:59.000Z

25

Management of Hanford Site non-defense production reactor spent nuclear fuel, Hanford Site, Richland, Washington  

SciTech Connect

The US Department of Energy (DOE) needs to provide radiologically, and industrially safe and cost-effective management of the non-defense production reactor spent nuclear fuel (SNF) at the Hanford Site. The proposed action would place the Hanford Site`s non-defense production reactor SNF in a radiologically- and industrially-safe, and passive storage condition pending final disposition. The proposed action would also reduce operational costs associated with storage of the non-defense production reactor SNF through consolidation of the SNF and through use of passive rather than active storage systems. Environmental, safety and health vulnerabilities associated with existing non-defense production reactor SNF storage facilities have been identified. DOE has determined that additional activities are required to consolidate non-defense production reactor SNF management activities at the Hanford Site, including cost-effective and safe interim storage, prior to final disposition, to enable deactivation of facilities where the SNF is now stored. Cost-effectiveness would be realized: through reduced operational costs associated with passive rather than active storage systems; removal of SNF from areas undergoing deactivation as part of the Hanford Site remediation effort; and eliminating the need to duplicate future transloading facilities at the 200 and 400 Areas. Radiologically- and industrially-safe storage would be enhanced through: (1) removal from aging facilities requiring substantial upgrades to continue safe storage; (2) utilization of passive rather than active storage systems for SNF; and (3) removal of SNF from some storage containers which have a limited remaining design life. No substantial increase in Hanford Site environmental impacts would be expected from the proposed action. Environmental impacts from postulated accident scenarios also were evaluated, and indicated that the risks associated with the proposed action would be small.

1997-03-01T23:59:59.000Z

26

Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel  

SciTech Connect

This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

1994-10-01T23:59:59.000Z

27

ENRICO FERMI FAST REACTOR SPENT NUCLEAR FUEL CRITICALLY CALCULATIONS: INTACT MODE  

SciTech Connect

The purpose of this calculation is to perform intact mode and partially degraded mode criticality evaluations of the Department of Energy's (DOE) Enrico Fermi (EF) Spent Nuclear Fuel (SNF) co-disposed in a 5 Defense High-Level Waste (5-DHLW) Waste Package (WP) and emplaced in a Monitored Geologic Repository (MGR). The criticality evaluations estimate the values of the effective neutron multiplication factor, k{sub eff}, a measure of nuclear criticality potential, for the 5-DHLW/DOE SNF WP with intact or partially degraded internal configurations. These evaluations contribute to the WP design.

A.S. Mobasheran

1999-04-12T23:59:59.000Z

28

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

SciTech Connect

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

D.R. Moscalu; L. Angers; J. Monroe-Rammsey; H.R. Radulesca

2000-07-21T23:59:59.000Z

29

Report to Congress on Plan for Interim Storage of Spent Nuclear...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from...

30

Developing and Qualifying Parameters for Closure Welding Overpacks Containing Research Reactor Spent Nuclear Fuel at Hanford  

SciTech Connect

Fluor engineers developed a Gas Tungsten Arc Welding (GTAW) technique and parameters, demonstrated requisite weld quality, and successfully closure-welded packaged spent nuclear fuel (SNF) overpacks at the Hanford Site. This paper reviews weld development and qualification activities associated with the overpack closure-welding and provides a summary of the production campaign. The primary requirement of the closure weld is to provide leak-tight confinement of the packaged material against release to the environment during interim storage (40-year design term). Required weld quality, in this case, was established through up-front development and qualification, and then verification of parameter compliance during production welding. This approach was implemented to allow for a simpler overpack design and more efficient production operations than possible with approaches using routine post-weld testing and nondestructive examination (NDE). A series of welding trials were conducted to establish the desired welding technique and parameters. Qualification of the process included statistical evaluation and American Society of Mechanical Engineers (ASME) Section IX testing. In addition, pull testing with a weighted mockup, and thermal calculation/physical testing to identify the maximum temperature the packaged contents would be subject to during welding, was performed. Thirteen overpacks were successfully packaged and placed into interim storage. The closure-welding development activities (including pull testing and thermal analysis) provided the needed confidence that the packaged SNF overpacks could be safely handled and placed into interim storage, and remain leak-tight for the duration of the storage term. (author)

Cannell, G.R.; Goldmann, L.H.; McCormack, R.L. [Hanford Site, Richland, WA (United States)

2008-07-01T23:59:59.000Z

31

DEVELOPING AND QUANTIFYING PARAMETERS FOR CLOSURE WELDING OVERPACKS CONTAINING RESEARCH REACTOR SPENT NUCLEAR FUEL AT HANFORD  

SciTech Connect

Fluor engineers developed a Gas Tungsten Arc Welding (GTAW) technique and parameters, demonstrated requisite weld quality and successfully closure-welded packaged spent nuclear fuel (SNF) overpacks at the Hanford Site. This paper reviews weld development and qualification activities associated with the overpack closure-welding and provides a summary of the production campaign. The primary requirement of the closure weld is to provide leaktight confinement of the packaged material against release to the environment during interim storage (40-year design term). Required weld quality, in this case, was established through up-front development and qualification, and then verification of parameter compliance during production welding. This approach was implemented to allow for a simpler overpack design and more efficient production operations than possible with approaches using routine post-weld testing and nondestructive examination (NDE). . A series of welding trials were conducted to establish the desired welding technique and parameters. Qualification of the process included statistical evaluation and American Society of Mechanical Engineers (ASME) Section IX testing. In addition, pull testing with a weighted mockup, and thermal calculation/physical testing to identify the maximum temperature the packaged contents would be subject to during welding, was performed. Thirteen overpacks were successfully packaged and placed into interim storage. The closure-welding development activities (including pull testing and thermal analysis) provided the needed confidence that the packaged SNF overpacks could be safely handled and placed into interim storage, and remain leaktight for the duration of the storage term.

CANNELL GR

2007-11-07T23:59:59.000Z

32

Cost Estimate for an Away-From-Reactor Generic Interim Storage Facility (GISF) for Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

As nuclear power plants began to run out of storage capacity in spent nuclear fuel (SNF) storage pools, many nuclear operating companies added higher density pool storage racks to increase pool capacity. Most nuclear power plant storage pools have been re-racked one or more times. As many spent fuel storage pools were re-racked to the maximum extent possible, nuclear operating companies began to employ interim dry storage technologies to store SNF in certified casks and canister-based systems outside of ...

2009-05-20T23:59:59.000Z

33

Nuclear Reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactors Nuclear reactors created not only large amounts of plutonium needed for the weapons programs, but a variety of other interesting and useful radioisotopes. They produced...

34

Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation  

SciTech Connect

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 spent 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.

Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

2008-07-01T23:59:59.000Z

35

EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

EIS-0203: Spent Nuclear Fuel Management and Idaho National EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs SUMMARY This EIS considers programmatic (DOE-wide) alternative approaches to safely, efficiently, and responsibly manage existing and projected quantities of spent nuclear fuel until the year 2035. This amount of time may be required to make and implement a decision on the ultimate disposition of spent nuclear fuel. DOE's spent nuclear fuel responsibilities include fuel generated by DOE production, research, and development reactors; naval reactors; university and foreign research reactors; domestic non-DOE reactors such as those at the National Institute

36

Report on interim storage of spent nuclear fuel  

SciTech Connect

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

1993-04-01T23:59:59.000Z

37

Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 3, Site team reports  

Science Conference Proceedings (OSTI)

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

Not Available

1993-11-01T23:59:59.000Z

38

NUCLEAR REACTOR  

DOE Patents (OSTI)

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

Treshow, M.

1961-09-01T23:59:59.000Z

39

Spent Nuclear Fuel Alternative Technology Decision Analysis  

SciTech Connect

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

Shedrow, C.B.

1999-11-29T23:59:59.000Z

40

Spent fuel utilization in a compact traveling wave reactor  

SciTech Connect

In recent years, several innovative designs of nuclear reactors are proposed. One of them is Traveling Wave Reactor (TWR). The unique characteristic of a TWR is the capability of breeding its own fuel in the reactor. The reactor is fueled by mostly depleted, natural uranium or spent nuclear fuel and a small amount of enriched uranium to initiate the fission process. Later on in the core, the reactor gradually converts the non-fissile material into the fissile in a process like a traveling wave. In this work, a TWR with spent nuclear fuel blanket was studied. Several parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, and fission power, were analyzed. The discharge burnup composition was also analyzed. The calculation is performed by a continuous energy Monte Carlo code McCARD.

Hartanto, Donny; Kim, Yonghee [Korea Advanced Institute of Science and Technology 373-1 Kusong-dong, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

2012-06-06T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

1962-10-23T23:59:59.000Z

42

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

E-Print Network (OSTI)

Fluorescence for Spent Nuclear Fuel Assay Brian J. Quiter ?of Pu isotopes in spent nuclear fuel (SNF). Given the lowU and 239 Pu in spent nuclear fuel using NRF. II. PERFORMING

Quiter, Brian

2012-01-01T23:59:59.000Z

43

Pyrochemical Treatment of Spent Nuclear Fuel  

SciTech Connect

Over the last 10 years, pyrochemical treatment of spent nuclear fuel has progressed from demonstration activities to engineering-scale production operations. As part of the Advanced Fuel Cycle Initiative within the U.S. Department of Energy’s Office of Nuclear Energy, Science and Technology, pyrochemical treatment operations are being performed as part of the treatment of fuel from the Experimental Breeder Reactor II at the Idaho National Laboratory. Integral to these treatment operations are research and development activities that are focused on scaling further the technology, developing and implementing process improvements, qualifying the resulting high-level waste forms, and demonstrating the overall pyrochemical fuel cycle.

K. M. Goff; K. L. Howden; G. M. Teske; T. A. Johnson

2005-10-01T23:59:59.000Z

44

NUCLEAR REACTOR  

DOE Patents (OSTI)

High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

Grebe, J.J.

1959-07-14T23:59:59.000Z

45

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described that includes spaced vertical fuel elements centrally disposed in a pressure vessel, a mass of graphite particles in the pressure vessel, means for fluidizing the graphite particles, and coolant tubes in the pressure vessel laterally spaced from the fuel elements. (AEC)

Post, R.G.

1963-05-01T23:59:59.000Z

46

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

Starr, C.

1963-01-01T23:59:59.000Z

47

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

Young, G.

1963-01-01T23:59:59.000Z

48

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

Christy, R.F.

1958-07-15T23:59:59.000Z

49

Nondestructive Spent Fuel Assay Using Nuclear Resonance Fluorescence  

E-Print Network (OSTI)

09-01188, ANS Advances in Nuclear Fuel Management IV, Hiltonanalysis of spent nuclear fuel via nuclear resonanceNondestructive Spent Fuel Assay Using Nuclear Resonance

Quiter, Brian

2010-01-01T23:59:59.000Z

50

Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 2, Working Group Assessment Team reports; Vulnerability development forms; Working group documents  

Science Conference Proceedings (OSTI)

The Secretary of Energy`s memorandum of August 19, 1993, established an initiative for a Department-wide assessment of the vulnerabilities of stored spent nuclear fuel and other reactor irradiated nuclear materials. A Project Plan to accomplish this study was issued on September 20, 1993 by US Department of Energy, Office of Environment, Health and Safety (EH) which established responsibilities for personnel essential to the study. The DOE Spent Fuel Working Group, which was formed for this purpose and produced the Project Plan, will manage the assessment and produce a report for the Secretary by November 20, 1993. This report was prepared by the Working Group Assessment Team assigned to the Hanford Site facilities. Results contained in this report will be reviewed, along with similar reports from all other selected DOE storage sites, by a working group review panel which will assemble the final summary report to the Secretary on spent nuclear fuel storage inventory and vulnerability.

Not Available

1993-11-01T23:59:59.000Z

51

Reactor-specific spent fuel discharge projections, 1984 to 2020  

Science Conference Proceedings (OSTI)

The original spent fuel utility data base (SFDB) has been adjusted to produce agreement with the EIA nuclear energy generation forecast. The procedure developed allows the detail of the utility data base to remain intact, while the overall nuclear generation is changed to match any uniform nuclear generation forecast. This procedure adjusts the weight of the reactor discharges as reported on the SFDB and makes a minimal (less than 10%) change in the original discharge exposures in order to preserve discharges of an integral number of fuel assemblies. The procedure used in developing the reactor-specific spent fuel discharge projections, as well as the resulting data bases themselves, are described in detail in this report. Discussions of the procedure cover the following topics: a description of the data base; data base adjustment procedures; addition of generic power reactors; and accuracy of the data base adjustments. Reactor-specific discharge and storage requirements are presented. Annual and cumulative discharge projections are provided. Annual and cumulative requirements for additional storage are shown for the maximum at-reactor (AR) storage assumption, and for the maximum AR with transshipment assumption. These compare directly to the storage requirements from the utility-supplied data, as reported in the Spent Fuel Storage Requirements Report. The results presented in this report include: the disaggregated spent fuel discharge projections; and disaggregated projections of requirements for additional spent fuel storage capacity prior to 1998. Descriptions of the methodology and the results are included in this report. Details supporting the discussions in the main body of the report, including descriptions of the capacity and fuel discharge projections, are included. 3 refs., 6 figs., 12 tabs.

Heeb, C.M.; Libby, R.A.; Holter, G.M.

1985-04-01T23:59:59.000Z

52

Nuclear Fuel Assembly and Related Methods for Spent Nuclear ...  

Nuclear Fuel Assembly and Related Methods for Spent Nuclear Fuel Reprocessing and Management Note: The technology described above is an early stage ...

53

Nuclear reactor  

DOE Patents (OSTI)

A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

Pennell, William E. (Greensburg, PA); Rowan, William J. (Monroeville, PA)

1977-01-01T23:59:59.000Z

54

Spent Nuclear Fuel (SNF) Project Execution Plan  

SciTech Connect

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.

LEROY, P.G.

2000-11-03T23:59:59.000Z

55

TEPP - Spent Nuclear Fuel | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

- Spent Nuclear Fuel - Spent Nuclear Fuel TEPP - Spent Nuclear Fuel This scenario provides the planning instructions, guidance, and evaluation forms necessary to conduct an exercise involving a highway shipment of spent nuclear fuel. This exercise manual is one in a series of five scenarios developed by the Department of Energy Transportation Emergency Preparedness Program. Responding agencies may include several or more of the following: local municipal and county fire, police, sheriff, and Emergency Medical Services (EMS) personnel; state, local, and federal emergency response teams; emergency response contractors;and other emergency response resources that could potentially be provided by the carrier and the originating facility (shipper). Spent Nuclear Fuel.docx More Documents & Publications

56

Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel  

E-Print Network (OSTI)

The ability to infer the reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3 percent, but only at high flux.

Hayes, A C

2012-01-01T23:59:59.000Z

57

Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel  

E-Print Network (OSTI)

The ability to infer the reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3 percent, but only at high flux.

A. C. Hayes; Gerard Jungman

2012-05-30T23:59:59.000Z

58

Spent Nuclear Fuel Transportation: An Overview  

Science Conference Proceedings (OSTI)

Spent nuclear fuel comprises a fraction of the hazardous materials packages shipped annually in the United States. In fact, at the present time, fewer than 100 packages of spent nuclear fuel are shipped annually. At the onset of spent fuel shipments to the proposed Yucca Mountain, Nevada, repository, the U.S. Department of Energy (DOE) expects to ship 400 - 500 spent fuel transport casks per year over the life of the facility. This study summarizes work on transportation cask design and testing, regulato...

2004-02-18T23:59:59.000Z

59

Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF)  

Energy.gov (U.S. Department of Energy (DOE))

GC-52 provides legal advice to DOE regarding the long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF). SNF is nuclear fuel that has been used as fuel in a reactor...

60

Transportation of Commercial Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

The U.S. industrys limited efforts at licensing transportation packages characterized as high-capacity, or containing high-burnup (>45 GWd/MTU) commercial spent nuclear fuel (CSNF), or both, have not been successful considering existing spent-fuel inventories that will have to be eventually transported. A holistic framework is proposed for resolving several CSNF transportation issues. The framework considers transportation risks, spent-fuel and cask-design features, and defense-in-depth in context of pre...

2010-12-10T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

NUCLEAR REACTOR  

DOE Patents (OSTI)

A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

Moore, R.V.; Bowen, J.H.; Dent, K.H.

1958-12-01T23:59:59.000Z

62

Nuclear Reactor Accidents  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactor Accidents The accidents at the Three Mile Island (TMI) and Chernobyl nuclear reactors have triggered particularly intense concern about radiation hazards. The TMI accident,...

63

Overview of the spent nuclear fuel project at Hanford  

SciTech Connect

The Spent Nuclear Fuel Project`s mission at Hanford is to {open_quotes}Provide safe, economic and environmentally sound management of Hanford spent nuclear fuel in a manner which stages it to final disposition.{close_quotes} The inventory of spent nuclear fuel (SNF) at the Hanford Site covers a wide variety of fuel types (production reactor to space reactor) in many facilities (reactor fuel basins to hot cells) at locations all over the Site. The 2,129 metric tons of Hanford SNF represents about 80% of the total US Department of Energy (DOE) inventory. About 98.5% of the Hanford SNF is 2,100 metric tons of metallic uranium production reactor fuel currently stored in the 1950s vintage K Basins in the 100 Area. This fuel has been slowly corroding, generating sludge and contaminating the basin water. This condition, coupled with aging facilities with seismic vulnerabilities, has been identified by several groups, including stakeholders, as being one of the most urgent safety and environmental concerns at the Hanford Site. As a direct result of these concerns, the Spent Nuclear Fuel Project was recently formed to address spent fuel issues at Hanford. The Project has developed the K Basins Path Forward to remove fuel from the basins and place it in dry interim storage. Alternatives that addressed the requirements were developed and analyzed. The result is a two-phased approach allowing the early removal of fuel from the K Basins followed by its stabilization and interim storage consistent with the national program.

Daily, J.L. [Dept. of Energy, Richland, WA (United States). Richland Operations Office; Fulton, J.C.; Gerber, E.W.; Culley, G.E. [Westinghouse Hanford Co., Richland, WA (United States)

1995-02-01T23:59:59.000Z

64

Air Shipment of Spent Nuclear Fuel from Romania to Russia  

SciTech Connect

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 for 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.

Igor Bolshinsky; Ken Allen; Lucian Biro; Alexander Buchelnikov

2010-10-01T23:59:59.000Z

65

Demonstration of a transportable storage system for spent nuclear fuel  

Science Conference Proceedings (OSTI)

The purpose of this paper is to discuss the joint demonstration project between the Sacramento Municipal Utility District (SMUD) and the US Department of Energy (DOE) regarding the use of a transportable storage system for the long-term storage and subsequent transport of spent nuclear fuel. SMUD's Rancho Seco nuclear generating station was shut down permanently in June 1989. After the shutdown, SMUD began planning the decommissioning process, including the disposition of the spent nuclear fuel. Concurrently, Congress had directed the Secretary of Energy to develop a plan for the use of dual-purpose casks. Licensing and demonstrating a dual-purpose cask, or transportable storage system, would be a step toward achieving Congress's goal of demonstrating a technology that can be used to minimize the handling of spent nuclear fuel from the time the fuel is permanently removed from the reactor through to its ultimate disposal at a DOE facility. For SMUD, using a transportable storage system at the Rancho Seco Independent Spent-Fuel Storage Installation supports the goal of abandoning Rancho Seco's spent-fuel pool as decommissioning proceeds.

Shetler, J.R.; Miller, K.R.; Jones, R.E. (Sacramento Municipal Utility District, Herald, CA (United States))

1993-01-01T23:59:59.000Z

66

NUCLEAR REACTOR  

DOE Patents (OSTI)

A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

Treshow, M.

1958-08-19T23:59:59.000Z

67

NUCLEAR REACTORS  

DOE Patents (OSTI)

An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

1961-12-01T23:59:59.000Z

68

Guidebook to nuclear reactors  

SciTech Connect

A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen.

Nero, A.V. Jr.

1976-05-01T23:59:59.000Z

69

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

SciTech Connect

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

Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki [Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 Nuclear Physics and Bio (Indonesia); Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 (Japan); Nuclear Physics and Bio Physics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia)

2012-06-06T23:59:59.000Z

70

Rack for storing spent nuclear fuel elements  

DOE Patents (OSTI)

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.

Rubinstein, Herbert J. (Los Gatos, CA); Clark, Philip M. (San Jose, CA); Gilcrest, James D. (San Jose, CA)

1978-06-20T23:59:59.000Z

71

Perils of plutonium [spent nuclear fuel storage  

Science Conference Proceedings (OSTI)

This paper focuses on the security of the ponds at reactor sites where radioactive spent fuel are being stored. A recent report by a panel of the National Academy of Sciences in Washington, DC, said that attacks by knowledgeable terrorists with access ...

P. P. Predd

2005-07-01T23:59:59.000Z

72

NUCLEAR REACTOR CONTROL SYSTEM  

DOE Patents (OSTI)

A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

1959-11-01T23:59:59.000Z

73

U.S. Spent Nuclear Fuel Data as of December 31, 2002  

Gasoline and Diesel Fuel Update (EIA)

Home > Nuclear > Spent Nuclear Fuel Home > Nuclear > Spent Nuclear Fuel Release Date: October 1, 2004 Next Release: Late 2010** Spent nuclear fuel data is collected by the Energy Information Administration (EIA) for the Office of Civilian Radioactive Waste Management (OCRWM). The spent nuclear fuel (SNF) data includes detailed characteristics of SNF generated by commercial U.S. nuclear power plants. From 1983 through 1995 this data was collected annually. Since 1996 this data has been collected every three years. The latest available detailed data covers all SNF discharged from commercial reactors before December 31, 2002, and is maintained in a data base by the EIA. Summary data tables from this data base may be found as indicated below. Table 1. Total U.S. Commercial Spent Nuclear Fuel Discharges, 1968 - 2002

74

Interim Storage of Used or Spent Nuclear Fuel Position Statement  

E-Print Network (OSTI)

The American Nuclear Society (ANS) supports the safe, controlled, licensed, and regulated interim storage of used nuclear fuel (UNF) (irradiated, spent fuel from a nuclear power reactor) until disposition can be determined and completed. ANS supports the U.S. Nuclear Regulatory Commission’s (NRC’s) determination that “spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 30 years beyond the licensed life for operation. ” 1 Current operational and decommissioned nuclear power plants in the United States were licensed with the expectation that the UNF would be stored at the nuclear power plant site until shipment to an interim storage facility, reprocessing plant, or permanent storage. Because of delays in Federal programs and policy issues, utilities have been forced to store UNF. Current means of interim storage of UNF at nuclear power plant sites include storage of discharged fuel in a water-filled pool or in a sealed dry cask, both under safe, controlled, and monitored conditions. This UNF interim storage is designed, managed, and controlled to minimize or preclude potential radiological hazards or material releases. At nuclear power plant sites in the United States and internationally, this interim storage is regulated under site license requirements and technical specifications imposed by the national or state regulator. In the United States, NRC is the licensing and regulatory authority. ANS believes that UNF interim storage

unknown authors

2008-01-01T23:59:59.000Z

75

Method for shearing spent nuclear fuel assemblies  

DOE Patents (OSTI)

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.

Weil, Bradley S. (Oak Ridge, TN); Watson, Clyde D. (Knoxville, TN)

1977-01-01T23:59:59.000Z

76

Description of the Canadian Particulate-Fill WastePackage (WP) System for Spent-Nuclear Fuel (SNF) and its Applicability to Ligh-Water Reactor SNF WPS with Depleted Uranium-Dioxide Fill  

NLE Websites -- All DOE Office Websites (Extended Search)

3502 3502 Chemical Technology Division DESCRIPTION OF THE CANADIAN PARTICULATE-FILL WASTE-PACKAGE (WP) SYSTEM FOR SPENT-NUCLEAR FUEL(SNF) AND ITS APPLICABILITY TO LIGHT- WATER REACTOR SNF WPS WITH DEPLETED URANIUM-DIOXIDE FILL Charles W. Forsberg Oak Ridge National Laboratory * P.O. Box 2008 Oak Ridge, Tennessee 37831-6180 Tel: (423) 574-6783 Fax: (423) 574-9512 Email: forsbergcw@ornl.gov October 20, 1997 _________________________ Managed by Lockheed Martin Energy Research Corp. under contract DE-AC05-96OR22464 for the * U.S. Department of Energy. iii CONTENTS LIST OF FIGURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v LIST OF TABLES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vii ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

77

Examining 239Pu and 240Pu Nuclear Resonance Fluorescence Measurements on Spent Fuel for Nuclear Safeguards  

E-Print Network (OSTI)

Fluorescence for Spent Nuclear Fuel Assay,” Inst. of Nucl.239 Pu content in spent nuclear fuel [4, 5]. Development ofin the context of spent nuclear fuel, summarizes the results

Quiter, Brian

2013-01-01T23:59:59.000Z

78

Apparatus for shearing spent nuclear fuel assemblies  

DOE Patents (OSTI)

A method and apparatus are described for shearing spent nuclear fuel assemblies of the type comprising an array of fuel pins disposed within an outer metal shell or shroud. A spent fuel assembly is first compacted in a known manner and then incrementally sheared using fixed and movable shear blades having matched laterally projecting teeth which slidably intermesh to provide the desired shearing action. Incremental advancement of the fuel assembly after each shear cycle is limited to a distance corresponding to the lateral projection of the teeth to ensure fuel assembly breakup into small uniform segments which are amenable to remote chemical processing.

Weil, Bradley S. (Knoxville, TN); Metz, III, Curtis F. (Knoxville, TN)

1980-01-01T23:59:59.000Z

79

Light Water Reactors Technology Development - Nuclear Reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Light Water Reactors Light Water Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

80

Heat Transfer Modeling of Dry Spent Nuclear Fuel Storage Facilities  

Science Conference Proceedings (OSTI)

The present work was undertaken to provide heat transfer model that accurately predicts the thermal performance of dry spent nuclear fuel storage facilities. One of the storage configurations being considered for DOE Aluminum-clad Spent Nuclear Fuel (Al-SNF), such as the Material and Testing Reactor (MTR) fuel, is in a dry storage facility. To support design studies of storage options a computational and experimental program has been conducted at the Savannah River Site (SRS). The main objective is to develop heat transfer models including natural convection effects internal to an interim dry storage canister and to geological codisposal Waste Package (WP). Calculated temperatures will be used to demonstrate engineering viability of a dry storage option in enclosed interim storage and geological repository WP and to assess the chemical and physical behaviors of the Al-SNF in the dry storage facilities. The current paper describes the modeling approaches and presents the computational results along with the experimental data.

Lee, S.Y.

1999-01-13T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

Spent nuclear fuel project integrated schedule plan  

SciTech Connect

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

Squires, K.G.

1995-03-06T23:59:59.000Z

82

Integrated process for reprocessing spent nuclear fuel  

DOE Patents (OSTI)

This invention is comprised of a process for recovering nuclear fuel from spent fuel assemblies that employs a single canister process container. The cladding and fuel are oxidized in the container, the fuel is dissolved and removed from the container for separation from the aqueous phase, the aqueous phase containing radioactive waste is returned to the container. This container is also the disposal vessel. Add solidification agents and compress container for long term storage.

Forsberg, C.W.

1991-03-06T23:59:59.000Z

83

Waste management plan for Hanford spent nuclear fuel characterization activities  

SciTech Connect

A joint project was initiated between Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratory (PNL) to address critical issues associated with the Spent Nuclear Fuel (SNF) stored at the Hanford Site. Recently, particular attention has been given to remediation of the SNF stored in the K Basins. A waste management plan (WMP) acceptable to both parties is required prior to the movement of selected material to the PNL facilities for examination. N Reactor and Single Pass Reactor (SPR) fuel has been stored for an extended period of time in the N Reactor, PUREX, K-East, and K-West Basins. Characterization plans call for transport of fuel material form the K Basins to the 327 Building Postirradiation Testing Laboratory (PTL) in the 300 Area for examination. However, PNL received a directive stating that no examination work will be started in PNL hot cell laboratories without an approved disposal route for all waste generated related to the activity. Thus, as part of the Characterization Program Management Plan for Hanford Spent Nuclear Fuel, a waste management plan which will ensure that wastes generated as a result of characterization activities conducted at PNL will be accepted by WHC for disposition is required. This document contains the details of the waste handling plan that utilizes, to the greatest extent possible, established waste handling and disposal practices at Hanford between PNL and WHC. Standard practices are sufficient to provides for disposal of most of the waste materials, however, special consideration must be given to the remnants of spent nuclear fuel elements following examination. Fuel element remnants will be repackaged in an acceptable container such as the single element canister and returned to the K Basins for storage.

Chastain, S.A. [Westinghouse Hanford Co., Richland, WA (United States); Spinks, R.L. [Pacific Northwest Lab., Richland, WA (United States)

1994-10-17T23:59:59.000Z

84

Nuclear reactor overflow line  

DOE Patents (OSTI)

The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

Severson, Wayne J. (Pittsburgh, PA)

1976-01-01T23:59:59.000Z

85

Remote Inspection Devices for Spent Reactor Enriched Uranium Fuel Elements  

SciTech Connect

A remote video inspection was developed and deployed in Argentina for the detailed inspection of highly radioactive spent reactor fuel (SNF) as a prerequisite to its shipment to the Savannah River Site (SRS) in the United States for long-term storage and disposition. The fuel is highly enriched uranium (HEU) spent assemblies dating from 1967 to 1989 and aluminum clad uranium-aluminum alloy of a typical material test reactor design. The specialized video system was designed for low cost, high portability, easy setup, and ease of usage, while accommodating the differing electrical systems (i.e. 110/60 Hz, 220/50 Hz) between the United States and Argentina.

Heckendorn, F.M.

2001-01-03T23:59:59.000Z

86

Nuclear reactor apparatus  

DOE Patents (OSTI)

A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.

Wade, Elman E. (Ruffs Dale, PA)

1978-01-01T23:59:59.000Z

87

NUCLEAR REACTORS AND EARTHQUAKES  

SciTech Connect

A book is presented which supplies pertinent seismological information to engineers in the nuclear reactor field. Data are presented on the occurrence, intensity, and wave shapes. Techniques are described for evaluating the response of structures to such events. Certain reactor types and their modes of operation are described briefly. Various protection systems are considered. Earthquake experience in industrial and reactor plants is described. (D.L.C.)

1961-01-01T23:59:59.000Z

88

Microsoft Word - spent nuclear fuel report.doc  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Management of Spent Nuclear Fuel Management of Spent Nuclear Fuel at the Savannah River Site DOE/IG-0727 May 2006 REPORT ON MANAGEMENT OF SPENT NUCLEAR FUEL AT THE SAVANNAH RIVER SITE TABLE OF CONTENTS Spent Nuclear Fuel Management Details of Finding 1 Recommendations 2 Comments 3 Appendices 1. Objective, Scope, and Methodology 4 2. Prior Audit Reports 5 3. Management Comments 6 SPENT NUCLEAR FUEL MANGEMENT Page 1 Details of Finding H-Canyon The Department of Energy's (Department) spent nuclear fuel Operations program at the Savannah River Site (Site) will likely require Extended H-Canyon to be maintained at least two years beyond defined operational needs. The Department committed to maintain H-Canyon operational readiness to provide a disposal path for

89

EIS-0279: Spent Nuclear Fuel Management, Aiken, South Carolina | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

79: Spent Nuclear Fuel Management, Aiken, South Carolina 79: Spent Nuclear Fuel Management, Aiken, South Carolina EIS-0279: Spent Nuclear Fuel Management, Aiken, South Carolina SUMMARY 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. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD April 5, 2013 EIS-0279: Amended Record of Decision Spent Nuclear Fuel Management at the Savannah River Site April 1, 2013 EIS-0279-SA-01: Supplement Analysis Savannah River Site Spent Nuclear Fuel Management (DOE/EIS-0279-SA-01 and

90

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel  

E-Print Network (OSTI)

09-01188, ANS Advances in Nuclear Fuel Management IV, HiltonParameter Library Spent Nuclear Fuel Transmission detector (Pu) mass in spent nuclear fuel (SNF) assemblies and to

Ludewigt, Bernhard A

2011-01-01T23:59:59.000Z

91

Anode Materials for Reprocessing of Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

In order to consume current stockpiles, uranium dioxide spent nuclear fuel will be .... and Synthesis of Intermetallic Clathrates for Energy Storage and Recovery.

92

Transportation capabilities study of DOE-owned spent nuclear fuel  

Science Conference Proceedings (OSTI)

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

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

1994-10-01T23:59:59.000Z

93

Spent Nuclear Fuel project integrated safety management plan  

SciTech Connect

This document is being revised in its entirety and the document title is being revised to ``Spent Nuclear Fuel Project Integrated Safety Management Plan.

Daschke, K.D.

1996-09-17T23:59:59.000Z

94

Record of Decision for the Savannah River Site Spent Nuclear...  

NLE Websites -- All DOE Office Websites (Extended Search)

Management Final Environmental Impact Statement, Aiken, SC AGENCY: Department of Energy (DOE). ACTION: Record of decision. SUMMARY: The Savannah River Site Spent Nuclear Fuel...

95

Pyroprocess for processing spent nuclear fuel  

DOE Patents (OSTI)

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

Miller, William E. (Naperville, IL); Tomczuk, Zygmunt (Lockport, IL)

2002-01-01T23:59:59.000Z

96

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel  

E-Print Network (OSTI)

and S.J. Thompson,“Determining Plutonium in Spent Fuel withTobin, “Determination of Plutonium Content in Spent FuelFluorescence to Measure Plutonium Mass in Spent Nuclear Fuel

Ludewigt, Bernhard A

2011-01-01T23:59:59.000Z

97

HOMOGENEOUS NUCLEAR POWER REACTOR  

DOE Patents (OSTI)

A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

King, L.D.P.

1959-09-01T23:59:59.000Z

98

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

SciTech Connect

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

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

2009-07-01T23:59:59.000Z

99

Examining 239Pu and 240Pu Nuclear Resonance Fluorescence Measurements on Spent Fuel for Nuclear Safeguards  

E-Print Network (OSTI)

S.D. Ambers, “Assesment of Nuclear Resonance Fluorescencefor Spent Nuclear Fuel Assay,” Inst. of Nucl. Mat. Man. ,clandestine material with nuclear resonance fluorescence,”

Quiter, Brian

2013-01-01T23:59:59.000Z

100

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - III: Spent DUPIC Fuel Disposal Cost  

Science Conference Proceedings (OSTI)

The disposal costs of spent pressurized water reactor (PWR), Canada deuterium uranium (CANDU) reactor, and DUPIC fuels have been estimated based on available literature data and the engineering design of a spent CANDU fuel disposal facility by the Atomic Energy of Canada Limited. The cost estimation was carried out by the normalization concept of total electricity generation. Therefore, the future electricity generation scale was analyzed to evaluate the appropriate capacity of the high-level waste disposal facility in Korea, which is a key parameter of the disposal cost estimation. Based on the total electricity generation scale, it is concluded that the disposal unit costs for spent CANDU natural uranium, CANDU-DUPIC, and PWR fuels are 192.3, 388.5, and 696.5 $/kg heavy element, respectively.

Ko, Won Il; Choi, Hangbok; Roh, Gyuhong; Yang, Myung Seung [Korea Atomic Energy Research Institute (Korea, Republic of)

2001-05-15T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

NUCLEAR BOMBS FROM LOW- ENRICHED URANIUM OR “SPENT ” FUEL  

E-Print Network (OSTI)

Conventional wisdom says that low-enriched uranium is not suitable for making nuclear weapons. However, an article in USA Today claims that “rogue ” states and terrorists have discovered that this is untrue. Not only that, but terrorists could separate plutonium from irradiated fuel (often called “spent fuel”) more easily than previously thought. (584.5495) WISE Amsterdam – Lowenriched uranium (LEU) is uranium containing up to 20 % uranium-235. Uranium with higher enrichment levels is classified as high-enriched, and is subject to international safeguards because it can be used to make nuclear weapons. However, a USA Today article claims that “rogue countries and terrorists” have discovered that it is possible to make nuclear weapons with uranium of lower enrichment, according to classified nuclear threat reports (1). The information is not entirely new. Back in 1996, a standard book on nuclear weapons material stated, “a self-sustaining chain reaction in a nuclear weapon cannot occur in depleted or natural or low-enriched uranium and is only theoretically IN THIS ISSUE: possible in LEU of roughly 10 percent or greater ” (2). Fuel for nuclear power reactors would not be suitable – this is typically enriched to 3-5 % uranium-235. However, for a “rogue state” wanting to make high-enriched uranium, it would take less work to start with nuclear fuel than with natural uranium. It could be done in a “small and easy to hide ” uranium enrichment plant – perhaps similar to the plant which has recently been discovered in Iran (3). Nevertheless, it would still require a substantial operation, since the fuel would need to be converted to uranium hexafluoride, enriched and then reconverted to uranium metal. More significantly, many research reactors use uranium of just under

unknown authors

2003-01-01T23:59:59.000Z

102

Risk and Responsibility Sharing in Nuclear Spent Fuel Management  

E-Print Network (OSTI)

With the Nuclear Waste Policy Act of 1982, the responsibility of American utilities in the long-term management of spent nuclear fuel was limited to the payment of a fee. This narrow involvement did not result in faster ...

De Roo, Guillaume

103

Nuclear reactor reflector  

DOE Patents (OSTI)

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

1994-01-01T23:59:59.000Z

104

Nuclear reactor reflector  

DOE Patents (OSTI)

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

Hopkins, R.J.; Land, J.T.; Misvel, M.C.

1994-06-07T23:59:59.000Z

105

Nuclear reactor control column  

DOE Patents (OSTI)

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

Bachovchin, Dennis M. (Plum Borough, PA)

1982-01-01T23:59:59.000Z

106

Nuclear & Uranium  

U.S. Energy Information Administration (EIA)

Nuclear & Uranium. Uranium fuel ... nuclear reactors, generation, spent fuel. Total Energy. Comprehensive data summaries, comparisons, analysis, and projections ...

107

Nano-particles for Spent Nuclear Fuel Separation  

Science Conference Proceedings (OSTI)

Symposium, Materials and Fuels for the Current and Advanced Nuclear Reactors III ... Development and Testing Advanced Ferritic Steels for Fast Reactor ...

108

Reactor & Nuclear Systems Publications | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Publications and Reports NSED Monthly Reports Reactor and Nuclear Systems Publications 2013 Publications 2012 Publications 2011 Publications 2010 and Older Publications Nuclear...

109

Nuclear Reactors and Technology  

SciTech Connect

This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

Cason, D.L.; Hicks, S.C. [eds.

1992-01-01T23:59:59.000Z

110

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

SciTech Connect

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

Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

2010-07-01T23:59:59.000Z

111

Radiological consequences of ship collisions that might occur in U.S. Ports during the shipment of foreign research reactor spent nuclear fuel to the United States in break-bulk freighters  

SciTech Connect

Accident source terms, source term probabilities, consequences, and risks are developed for ship collisions that might occur in U.S. ports during the shipment of spent fuel from foreign research reactors to the United States in break-bulk freighters.

Sprung, J.L.; Bespalko, S.J.; Massey, C.D.; Yoshimura, R. [Sandia National Laboratory, Albuquerque, NM (United States); Johnson, J.D. [GRAM Inc., Albuquerque, NM (United States); Reardon, P.C. [PCRT Technologies, Albuquerque, NM (United States); Ebert, M.W.; Gallagher D.W. [Science Applications International Corp., Reston, VA (United States)

1996-08-01T23:59:59.000Z

112

U.S. Spent Nuclear Fuel Data as of December 31, 1998  

Gasoline and Diesel Fuel Update (EIA)

Spent Nuclear Fuel Data, Detailed United States as of December 31, 1998 Spent nuclear fuel data is collected by the Energy Information Administration (EIA) for the Office of Civilian Radioactive Waste Management (OCRWM). The spent nuclear fuel (SNF) data includes detailed characteristics of SNF generated by commercial U.S. nuclear power plants. From 1983 through 1995 this data was collected annually. Since 1996 this data has been collected every three years. The latest available detailed data covers all SNF discharged from commercial reactors before December 31, 1998, and is maintained in a database by the EIA. Summary data tables from this database may be found as indicated below. The detailed data are available on request from Jim Finucane who can be reached at 202-287-1966 or at

113

THERMAL NUCLEAR REACTOR  

DOE Patents (OSTI)

Nuclear reactors of the graphite moderated air cooled type in which canned slugs or rods of fissile material are employed are discussed. Such a reactor may be provided with a means for detecting dust particles in the exhausted air. The means employed are lengths of dust absorbent cord suspended in vertical holes in the shielding structure above each vertical coolant flow channel to hang in the path of the cooling air issuing from the channels, and associated spindles and drive motors for hauling the cords past detectors, such as Geiger counters, for inspecting the cords periodically. This design also enables detecting the individual channel in which a fault condition may have occurred.

Fenning, F.W.; Jackson, R.F.

1957-09-24T23:59:59.000Z

114

SUPPLEMENT ANALYSIS OF FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL TRANSPORTATION ALONG OTHER THAN~. PRESENTATIVE ROUTE FROM CONCORD NAVAL WEAPO~~ STATION TO IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LADORA TORY Introduction The Department of Energy is planning to transport foreign research reactor spent nuclear fuel by rail from the Concord Naval Weapons Station (CNWS), Concord, California, to the Idaho National Engineering and Environmental Laboratory (INEEL). The environmental analysis supporting the decision to transport, by rail or truck, foreign research reactor spent nuclear fuel from CNWS to the INEEL is contained in +he Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliftration Policy Concerning Foreign Research Reactor

115

Spent Nuclear Fuel Project operational staffing plan  

SciTech Connect

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

Debban, B.L.

1996-03-01T23:59:59.000Z

116

DOE-owned spent nuclear fuel strategic plan. Revision 1  

SciTech Connect

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

1996-09-01T23:59:59.000Z

117

Test storage of spent reactor fuel in the Climax granite at the Nevada Test Site  

SciTech Connect

A test of retrievable dry geologic storage of spent fuel assemblies from an operating commercial nuclear reactor is underway at the Nevada Test Site. This generic test is located 420 m below the surface in the Climax granitic stock. Eleven canisters of spent fuel approximately 2.3 years out of reactor core (about 2 kW/canister thermal output) will be emplaced in a storage drift along with 6 electrical simulator canisters and their effects will be compared. Two adjacent drifts will contain electrical heaters, which will be operated to simulate within the test array the thermal field of a large repository. The test objectives, technical concepts and rationale, and details of the test are stated and discussed.

Ramspott, L.D.; Ballou, L.B.

1980-02-13T23:59:59.000Z

118

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

SciTech Connect

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

Not Available

1994-06-01T23:59:59.000Z

119

Nuclear reactor safety device  

DOE Patents (OSTI)

A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

Hutter, Ernest (Wilmette, IL)

1986-01-01T23:59:59.000Z

120

Heat dissipating nuclear reactor  

DOE Patents (OSTI)

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

1987-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Heat dissipating nuclear reactor  

DOE Patents (OSTI)

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

Hunsbedt, A.; Lazarus, J.D.

1985-11-21T23:59:59.000Z

122

Advanced Reactor Development and Technology - Nuclear Engineering...  

NLE Websites -- All DOE Office Websites (Extended Search)

Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Nuclear Data Program Advanced Reactor Development Overview Advanced Fast Reactor...

123

Locations of spent nuclear fuel and high-level radioactive waste ultimately destined for geologic disposal  

Science Conference Proceedings (OSTI)

Since the late 1950s, Americans have come to rely more and more on energy generated from nuclear reactors. Today, 109 commercial nuclear reactors supply over one-fifth of the electricity used to run our homes, schools, factories, and farms. When the nuclear fuel can no longer sustain a fission reaction in these reactors it becomes `spent` or `used` and is removed from the reactors and stored onsite. Most of our Nation`s spent nuclear fuel is currently being stored in specially designed deep pools of water at reactor sites; some is being stored aboveground in heavy thick-walled metal or concrete structures. Sites currently using aboveground dry storage systems include Virginia Power`s Surry Plant, Carolina Power and Light`s H.B. Robinson Plant, Duke Power`s Oconee Nuclear Station, Colorado Public Service Company`s shutdown reactor at Fort St. Vrain, Baltimore Gas and Electric`s Calvert Cliffs Plant, and Michigan`s Consumer Power Palisades Plant.

Not Available

1994-09-01T23:59:59.000Z

124

Experience in Using Fills for Spent Nuclear Fuel Waste Packages  

NLE Websites -- All DOE Office Websites (Extended Search)

Fills for SNF Waste Packages Experience in Using Fills for Spent Nuclear Fuel Waste Packages The use of other fill materials in waste packages has been investigated by several...

125

EIS-0250: Geologic Repository for the Disposal of Spent Nuclear...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

EIS-0250: Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada EIS-0250: Geologic Repository for the...

126

Handbook on Neutron Absorber Materials for Spent Nuclear Fuel Applications  

Science Conference Proceedings (OSTI)

This handbook is intended to become a single source of information regarding technical characteristics of neutron absorber materials that have been used for storage and transportation of spent nuclear fuel as well as to provide a summary of users' experience.

2005-12-08T23:59:59.000Z

127

Safety Aspects of Wet Storage of Spent Nuclear Fuel, OAS-L-13-11  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Safety Aspects of Wet Storage of Safety Aspects of Wet Storage of Spent Nuclear Fuel OAS-L-13-11 July 2013 Department of Energy Washington, DC 20585 July 10, 2013 MEMORANDUM FOR THE SENIOR ADVISOR FOR ENVIRONMENTAL MANAGEMENT FROM: Daniel M. Weeber Assistant Inspector General for Audits and Administration Office of Inspector General SUBJECT: INFORMATION: Audit Report on "Safety Aspects of Wet Storage of Spent Nuclear Fuel" BACKGROUND The Department of Energy (Department) is responsible for managing and storing spent nuclear fuel (SNF) generated by weapons and research programs and recovered through nonproliferation programs. The SNF consists of irradiated reactor fuel and cut up assemblies containing uranium, thorium and/or plutonium. The Department stores 34 metric tons of heavy metal SNF primarily

128

Proliferation Resistant Nuclear Reactor Fuel  

Science Conference Proceedings (OSTI)

Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and we posit that the exploration, development, and implementation of intrinsic mechanisms such as discussed here are part of a balanced approach aimed at preventing the misuse of nuclear material for nuclear-energy applications.

Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

2011-02-18T23:59:59.000Z

129

Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement  

SciTech Connect

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

N /A

2000-04-14T23:59:59.000Z

130

Nuclear Reactor Technologies | Department of Energy  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactor Technologies Nuclear Reactor Technologies TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority TVA Watts Bar Nuclear Power Plant | Photo...

131

Nuclear reactor building  

DOE Patents (OSTI)

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

Gou, P.F.; Townsend, H.E.; Barbanti, G.

1994-04-05T23:59:59.000Z

132

Nuclear reactor building  

DOE Patents (OSTI)

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

Gou, Perng-Fei (Saratoga, CA); Townsend, Harold E. (Campbell, CA); Barbanti, Giancarlo (Sirtori, IT)

1994-01-01T23:59:59.000Z

133

Nuclear reactor safety device  

DOE Patents (OSTI)

A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

Hutter, E.

1983-08-15T23:59:59.000Z

134

HOMOGENEOUS NUCLEAR REACTOR  

DOE Patents (OSTI)

Nuclear reactors of the homogeneous liquid fuel type are discussed. The reactor is comprised of an elongated closed vessel, vertically oriented, having a critical region at the bottom, a lower chimney structure extending from the critical region vertically upwardly and surrounded by heat exchanger coils, to a baffle region above which is located an upper chimney structure containing a catalyst functioning to recombine radiolyticallydissociated moderator gages. In operation the liquid fuel circulates solely by convection from the critical region upwardly through the lower chimney and then downwardly through the heat exchanger to return to the critical region. The gases formed by radiolytic- dissociation of the moderator are carried upwardly with the circulating liquid fuel and past the baffle into the region of the upper chimney where they are recombined by the catalyst and condensed, thence returning through the heat exchanger to the critical region.

Hammond, R.P.; Busey, H.M.

1959-02-17T23:59:59.000Z

135

Nuclear reactor control apparatus  

DOE Patents (OSTI)

Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

Sridhar, Bettadapur N. (Cupertino, CA)

1983-11-01T23:59:59.000Z

136

The U.S.-Russian joint studies on using power reactors to disposition surplus weapon plutonium as spent fuel  

SciTech Connect

In 1996, the US and the Russian Federation completed an initial joint study of the candidate options for the disposition of surplus weapons plutonium in both countries. The options included long term storage, immobilization of the plutonium in glass or ceramic for geologic disposal, and the conversion of weapons plutonium to spent fuel in power reactors. For the latter option, the US is only considering the use of existing light water reactors (LWRs) with no new reactor construction for plutonium disposition, or the use of Canadian deuterium uranium (CANDU) heavy water reactors. While Russia advocates building new reactors, the cost is high, and the continuing joint study of the Russian options is considering only the use of existing VVER-1000 LWRs in Russia and possibly Ukraine, the existing BN-60O fast neutron reactor at the Beloyarsk Nuclear Power Plant in Russia, or the use of the Canadian CANDU reactors. Six of the seven existing VVER-1000 reactors in Russia and the eleven VVER-1000 reactors in Ukraine are all of recent vintage and can be converted to use partial MOX cores. These existing VVER-1000 reactors are capable of converting almost 300 kg of surplus weapons plutonium to spent fuel each year with minimum nuclear power plant modifications. Higher core loads may be achievable in future years.

Chebeskov, A.; Kalashnikov, A. [State Scientific Center, Obninsk (Russian Federation). Inst. of Physics and Power Engineering; Bevard, B.; Moses, D. [Oak Ridge National Lab., TN (United States); Pavlovichev, A. [State Scientific Center, Moscow (Russian Federation). Kurchatov Inst.

1997-09-01T23:59:59.000Z

137

Nuclear Reactor Severe Accident Experiments  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.

138

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

SciTech Connect

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

NONE

1995-09-01T23:59:59.000Z

139

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

SciTech Connect

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

Quiter, Brian; Ludewigt, Bernhard; Ambers, Scott

2011-06-30T23:59:59.000Z

140

Separator assembly for use in spent nuclear fuel shipping cask  

DOE Patents (OSTI)

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.

Bucholz, James A. (Oak Ridge, TN)

1983-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Managing Spent Nuclear Fuel at the Idaho National Laboratory  

SciTech Connect

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

Thomas Hill; Denzel L. Fillmore

2005-10-01T23:59:59.000Z

142

Spent nuclear fuel Canister Storage Building CDR Review Committee report  

SciTech Connect

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

Dana, W.P.

1995-12-01T23:59:59.000Z

143

Cold Demonstration of a Spent Nuclear Fuel Dry Transfer System  

Science Conference Proceedings (OSTI)

The development of a spent nuclear fuel dry transfer system (DTS) has moved from the design phase to demonstration of major components. Use of an on-site DTS allows utilities with limited crane capacities or other plant restrictions to take advantage of large efficient storage systems. This system also permits utilities to transfer spent fuel from loaded storage casks to transport casks without returning to their fuel storage pool, a circumstance that may arise during the decommissioning process.

1999-09-24T23:59:59.000Z

144

Subcritical transmutation of spent nuclear fuel.  

E-Print Network (OSTI)

??A series of fuel cycle simulations were performed using CEA's reactor physics code ERANOS 2.0 to analyze the transmutation performance of the Subcritical Advanced Burner… (more)

Sommer, Christopher Michael

2011-01-01T23:59:59.000Z

145

Overview of Reactor and Nuclear  

E-Print Network (OSTI)

and Safety Gary Mays Nuclear Data and Criticality Safety Mike Dunn Nuclear Security Modeling Tim Valentine - Office of Environmental Management - Office of Intelligence · National Nuclear Security AdministrationOverview of Reactor and Nuclear Systems Division Cecil Parks RNS Division Director parkscv

146

EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

An emergency shutdown or scram apparatus for use in a nuclear reactor that includes a neutron absorber suspended from a temperature responsive substance that is selected to fail at a preselected temperature in excess of the normal reactor operating temperature, whereby the neutron absorber is released and allowed to fall under gravity to a preselected position within the reactor core is presented. (AEC)

Paget, J.A.; Koutz, S.L.; Stone, R.S.; Stewart, H.B.

1963-12-24T23:59:59.000Z

147

SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES  

SciTech Connect

Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier, initial {sup 235}U enrichment, and time of discharge from the reactor as well as the assigned burnup, but the distribution. of burnup axially along the assembly length is not provided. The axial burnup profile is maintained within acceptable bounds by the operating conditions of the nuclear reactor and is calculated during preparations to reload a reactor, but the actual burnup profile is not measured. The axial burnup profile is important to the determination of the reactivity of a waste package, so a conservative evaluation of the calculated axial profiles for a large database of SNF has been performed. The product of the axial profile evaluation is a profile that is conservative. Thus, there is no need for physical measurement of the axial profile. The assembly identifier is legible on each SNF assembly and the utility records provide the associated characteristics of the assembly. The conservative methodologies used to determine the criticality loading curve for a waste package provide sufficient margin so that criticality safety is assured for preclosure operations even in the event of a misload. Consideration of misload effects for postclosure time periods is provided by the criticality Features, Events, and Processes (FEPs) analysis. The conservative approaches used to develop and apply the criticality loading curve are thus sufficiently robust that the utility assigned burnup is an adequate source of burnup values, and additional means of verification of assigned burnup through physical measurements are not needed.

BSC

2004-12-01T23:59:59.000Z

148

Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania  

SciTech Connect

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

K. J. Allen; I. Bolshinsky; L. L. Biro; M. E. Budu; N. V. Zamfir; M. Dragusin

2010-07-01T23:59:59.000Z

149

Nuclear Regulatory Commission's Integrated Strategy for Spent Fuel Management  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

NRC's NRC's Integrated Strategy for NRC s Integrated Strategy for Spent Fuel Management Earl Easton 1 U.S. Nuclear Regulatory Commission May 25, 2010 Road to Yucca Mountain * 20+ years of preparation for the licensing i review * DOE application received in June 2008 and accepted for review in September 2008 * President Obama pursues alternatives to Yucca Mountain * DOE motion to withdraw in March 2010 2 * DOE motion to withdraw in March 2010 * Blue Ribbon Commission on America's Nuclear Future 2 Growing Spent Fuel Inventory Cumulative Used Nuclear Fuel Scenarios 50,000 100,000 150,000 200,000 250,000 Metric Tons 3 - 50,000 2010 2015 2020 2025 2030 2035 2040 2045 2050 Year Reference: Crozat, March 2010 Integrated Strategy * In response to the evolving national debate on spent fuel management strategy, NRC initiated a number of actions:

150

Nuclear reactor control apparatus  

DOE Patents (OSTI)

Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod.

Sridhar, Bettadapur N. (Cupertino, CA)

1983-10-25T23:59:59.000Z

151

GAS COOLED NUCLEAR REACTORS  

DOE Patents (OSTI)

A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

Long, E.; Rodwell, W.

1958-06-10T23:59:59.000Z

152

Nuclear reactor control  

DOE Patents (OSTI)

1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

Cawley, William E. (Phoenix, AZ); Warnick, Robert F. (Pasco, WA)

1982-01-01T23:59:59.000Z

153

Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography  

E-Print Network (OSTI)

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

Jonkmans, G; Jewett, C; Thompson, M

2012-01-01T23:59:59.000Z

154

Analysis of the risk of transporting spent nuclear fuel by train  

SciTech Connect

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

Elder, H.K.

1981-09-01T23:59:59.000Z

155

Application of Neutron-Absorbing Structural-Amorphous metal (SAM) Coatings for Spent Nuclear Fuel (SNF) Container to Enhance Criticality Safety Controls  

E-Print Network (OSTI)

Metal Coatings for Spent Nuclear Fuel (SNF) Containers: UseCoatings for Spent Nuclear Fuel (SNF) Container to Enhance2006 ABSTRACT Spent nuclear fuel contains fissionable

2006-01-01T23:59:59.000Z

156

Fast Reactor Curriculum Workshop - Nuclear Engineering Division...  

NLE Websites -- All DOE Office Websites (Extended Search)

Fast Reactor Curriculum Workshop Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear...

157

Assessment of the risk of transporting spent nuclear fuel by truck  

SciTech Connect

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

Elder, H.K.

1978-11-01T23:59:59.000Z

158

Criticality Risks During Transportation of Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

This report presents a best-estimate probabilistic risk assessment (PRA) to quantify the frequency of criticality accidents during railroad transportation of spent nuclear fuel casks. The assessment is of sufficient detail to enable full scrutiny of the model logic and the basis for each quantitative parameter contributing to criticality accident scenario frequencies.

2006-12-14T23:59:59.000Z

159

Railroad transportation of spent nuclear fuel  

Science Conference Proceedings (OSTI)

This report documents a detailed analysis of rail operations that are important for assessing the risk of transporting high-level nuclear waste. The major emphasis of the discussion is towards ''general freight'' shipments of radioactive material. The purpose of this document is to provide a basis for selecting models and parameters that are appropriate for assessing the risk of rail transportation of nuclear waste.

Wooden, D.G.

1986-03-01T23:59:59.000Z

160

Spent nuclear fuel project detonation phenomena of hydrogen/oxygen in spent fuel containers  

DOE Green Energy (OSTI)

Movement of Spent N Reactor fuels from the Hanford K Basins near the Columbia River to Dry interim storage facility on the Hanford plateau will require repackaging the fuel in the basins into multi-canister overpacks (MCOs), drying of the fuel, transporting the contained fuel, hot conditioning, and finally interim storage. Each of these functions will be accomplished while the fuel is contained in the MCOs by several mechanisms. The principal source of hydrogenand oxygen within the MCOs is residual water from the vacuum drying and hot conditioning operations. This document assesses the detonation phenomena of hydrogen and oxygen in the spent fuel containers. Several process scenarios have been identified that could generate detonation pressures that exceed the nominal 10 atmosphere design limit ofthe MCOS. Only 42 grams of radiolized water are required to establish this condition.

Cooper, T.D.

1996-09-30T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

K Basin spent nuclear fuel characterization  

SciTech Connect

The results of the characterization efforts completed for the N Reactor fuel stored in the Hanford K Basins were Collected and summarized in this single referencable document. This summary provides a ''road map'' for what was done and the results obtained for the fuel characterization program initiated in 1994 and scheduled for completion in 1999 with the fuel oxidation rate measurement under moist inert atmospheres.

LAWRENCE, L.A.

1999-02-10T23:59:59.000Z

162

Issues relating to spent nuclear fuel storage on the Oak Ridge Reservation  

Science Conference Proceedings (OSTI)

Currently, about 2,800 metric tons of spent nuclear fuel (SNF) is stored in the US, 1,000 kg of SNF (or about 0.03% of the nation`s total) are stored at the US Department of Energy (DOE) complex in Oak Ridge, Tennessee. However small the total quantity of material stored at Oak Ridge, some of the material is quite singular in character and, thus, poses unique management concerns. The various types of SNF stored at Oak Ridge will be discussed including: (1) High-Flux Isotope Reactor (HFIR) and future Advanced Neutron Source (ANS) fuels; (2) Material Testing Reactor (MTR) fuels, including Bulk Shielding Reactor (BSR) and Oak Ridge Research Reactor (ORR) fuels; (3) Molten Salt Reactor Experiment (MSRE) fuel; (4) Homogeneous Reactor Experiment (HRE) fuel; (5) Miscellaneous SNF stored in Oak Ridge National Laboratory`s (ORNL`s) Solid Waste Storage Areas (SWSAs); (6) SNF stored in the Y-12 Plant 9720-5 Warehouse including Health. Physics Reactor (HPRR), Space Nuclear Auxiliary Power (SNAP-) 10A, and DOE Demonstration Reactor fuels.

Klein, J.A.; Turner, D.W.

1994-12-31T23:59:59.000Z

163

Nuclear Reactors and Technology; (USA)  

SciTech Connect

Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

Cason, D.L.; Hicks, S.C. (eds.)

1991-01-01T23:59:59.000Z

164

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network (OSTI)

Fundamental aspects of nuclear reactor fuel elements.Unlike permanent nuclear reactor core components, nuclearof the first nuclear reactors, commercial nuclear fuel still

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

165

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

E-Print Network (OSTI)

the National Nuclear Security Administration, US Departmentof Energys National Nuclear Security Administration (NNSA)

Quiter, Brian

2012-01-01T23:59:59.000Z

166

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

E-Print Network (OSTI)

Security of the National Nuclear Security Administration, USof Energys National Nuclear Security Administration (NNSA)

Quiter, Brian

2012-01-01T23:59:59.000Z

167

Locations of Spent Nuclear Fuel and High-Level Radioactive Waste  

Energy.gov (U.S. Department of Energy (DOE))

Map of the United States of America showing the locations of spent nuclear fuel and high-level radioactive waste.

168

Nuclear reactor I  

DOE Patents (OSTI)

A nuclear reactor, particularly a liquid-metal breeder reactor whose upper internals include provision for channeling the liquid metal flowing from the core-component assemblies to the outlet plenum in vertical paths in direction generally along the direction of the respective assemblies. The metal is channeled by chimneys, each secured to, and extending from, a grid through whose openings the metal emitted by a plurality of core-component assemblies encompassed by the grid flows. To reduce the stresses resulting from structural interaction, or the transmissive of thermal strains due to large temperature differences in the liquid metal emitted from neighboring core-component assemblies, throughout the chimneys and the other components of the upper internals, the grids and the chimneys are supported from the heat plate and the core barrel by support columns (double portal support) which are secured to the head plate at the top and to a member, which supports the grids and is keyed to the core barrel, at the bottom. In addition to being restrained from lateral flow by the chimneys, the liquid metal is also restrained from flowing laterally by a peripheral seal around the top of the core. This seal limits the flow rate of liquid metal, which may be sharply cooled during a scram, to the outlet nozzles. The chimneys and the grids are formed of a highly-refractory, high corrosion-resistant nickel-chromium-iron alloy which can withstand the stresses produced by temperature differences in the liquid metal. The chimneys are supported by pairs of plates, each pair held together by hollow stubs coaxial with, and encircling, the chimneys. The plates and stubs are a welded structure but, in the interest of economy, are composed of stainless steel which is not weld compatible with the refractory metal. The chimneys and stubs are secured together by shells of another nickel-chromium-iron alloy which is weld compatible with, and is welded to, the stubs and has about the same coefficient of expansion as the highly-refractory, high corrosion-resistant alloy.

Ference, Edward W. (Central City, PA); Houtman, John L. (Acme, PA); Waldby, Robert N. (New Stanton, PA)

1977-01-01T23:59:59.000Z

169

Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor  

Science Conference Proceedings (OSTI)

This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

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

2011-01-01T23:59:59.000Z

170

Advanced Nuclear Reactors | Department of Energy  

NLE Websites -- All DOE Office Websites (Extended Search)

in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor...

171

The Properties of Spent Nuclear Fuel under Waste Disposal ...  

Science Conference Proceedings (OSTI)

Symposium, Materials Issues in Nuclear Waste Management in the 21st Century ... UO2 in the form of a ceramic pellet with a density close to theoretical. ... On discharge fro reactor the pellets have undergone a number of physical and ...

172

Advanced Nuclear Research Reactor  

SciTech Connect

This report describes technical modifications implemented by INVAP to improve the safety of the Research Reactors the company designs and builds.

Lolich, J.V.

2004-10-06T23:59:59.000Z

173

Spent Nuclear Fuel Self-Induced XRF to Predict Pu to U Content  

E-Print Network (OSTI)

The quantification of plutonium (Pu) in spent nuclear fuel is an increasingly important safeguards issue. There exists an estimated worldwide 980 metric tons of Pu in the nuclear fuel cycle and the majority is in spent nuclear fuel waiting for long term storage or fuel reprocessing. This study investigates utilizing the measurement of x-ray fluorescence (XRF) from the spent fuel for the quantification of its uranium (U) to Pu ratio. Pu quantification measurements at the front end of the reprocessing plant, the fuel cycle area of interest, would improve input accountability and shipper/receiver differences. XRF measurements were made on individual PWR fuel rods with varying fuel ages and final burn-ups at Oak Ridge National Laboratory (ORNL) in July 2008 and January 2009. These measurements successfully showed that it is possible to measure the Pu x-ray peak at 103.7 keV in PWR spent fuel (~1 percent Pu) using a planar HPGe detector. Prior to these measurement campaigns, the Pu peak has only been measured for fast breeder reactor fuel (~40 percent Pu). To understand the physics of the measurements, several modern physics simulations were conducted to determine the fuel isotopics, the sources of XRF in the spent fuel, and the sources of Compton continuum. Fuel transformation and decay simulations demonstrated the Pu/U measured peak ratio is directly proportional to the Pu/U content and increases linearly as burn-up increases. Spent fuel source simulations showed for 4 to 13 year old PWR fuel with burn-up ranges from 50 to 67 GWd/MTU, initial photon sources and resulting Compton and XRF interactions adequately model the spent fuel measured spectrum and background. The detector simulations also showed the contributions to the Compton continuum from strongest to weakest are as follows: the fuel, the shipping tube, the cladding, the detector can, the detector crystal and the collimator end. The detector simulations showed the relationship between the Pu/U peak ratio and fuel burn-up over predict the measured Pu/U peak but the trend is the same. In conclusion, the spent fuel simulations using modern radiation transport physics codes can model the actual spent fuel measurements but need to be benchmarked.

Stafford, Alissa Sarah

2010-08-01T23:59:59.000Z

174

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

E-Print Network (OSTI)

of the Institute of Nuclear Material Management, Tucson, AZ,Assay, Institute of Nuclear Materials Management 51st Annual

Quiter, Brian

2012-01-01T23:59:59.000Z

175

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

SciTech Connect

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

NONE

1995-09-01T23:59:59.000Z

176

THE ECONOMICS OF REPROCESSING vs DIRECT DISPOSAL OF SPENT NUCLEAR FUEL  

SciTech Connect

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

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

2003-07-01T23:59:59.000Z

177

Uncanistered Spent Nuclear fuel Disposal Container System Description Document  

Science Conference Proceedings (OSTI)

The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in the emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Multiple boiling water reactor (BWR) and pressurized water reactor (PWR) disposal container designs are needed to accommodate the expected range of spent fuel assemblies and provide long-term confinement of the commercial SNF. The disposal container will include outer and inner cylinder walls, outer cylinder lids (two on the top, one on the bottom), inner cylinder lids (one on the top, one on the bottom), and an internal metallic basket structure. Exterior labels will provide a means by which to identify the disposal container and its contents. The two metal cylinders, in combination with the cladding, Emplacement Drift System, drip shield, and natural barrier, will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel and the outer cylinder and outer cylinder lid will be made of high-nickel alloy. The basket will assist criticality control, provide structural support, and improve heat transfer. The Uncanistered SNF Disposal Container System interfaces with the emplacement drift environment and internal waste by transferring heat from the SNF to the external environment and by protecting the SFN assemblies and their contents from damage/degradation by the external environment. The system also interfaces with the SFN by limiting access of moderator and oxidizing agents of the SFN. The waste package interfaces with the Emplacement Drift System's emplacement drift pallets upon which the wasted packages are placed. The disposal container interfaces with the Assembly Transfer System, Waste Emplacement/Retrieval System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement and retrieval of the disposal container/waste package.

NONE

2000-10-12T23:59:59.000Z

178

NUCLEAR REACTOR FUEL SYSTEMS  

DOE Patents (OSTI)

Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

1959-09-15T23:59:59.000Z

179

Molten tin reprocessing of spent nuclear fuel elements  

DOE Patents (OSTI)

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

Heckman, Richard A. (Castro Valley, CA)

1983-01-01T23:59:59.000Z

180

Naval Spent Nuclear Fuel disposal Container System Description Document  

Science Conference Proceedings (OSTI)

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

N. E. Pettit

2001-07-13T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

Criticality Risks During Transportation of Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

This report presents a best-estimate probabilistic risk assessment (PRA) to quantify the frequency of criticality accidents during railroad transportation of spent nuclear fuel casks. The assessment is of sufficient detail to enable full scrutiny of the model logic and the basis for each quantitative parameter contributing to criticality accident scenario frequencies. The report takes into account the results of a 2007 peer review of the initial version of this probabilistic risk assessment, which was pu...

2008-12-10T23:59:59.000Z

182

Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements  

SciTech Connect

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

KLEM, M.J.

2000-10-18T23:59:59.000Z

183

Fast Reactor Spent Fuel Processing: Experience and Criticality Safety  

SciTech Connect

This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day operations as well as obtaining historical information. Over 12,000 driver fuel elements have been processed resulting in the production of 2500 kg of 20% enriched uranium. Also, over one thousand blanket fuel elements have been processed resulting in the production of 2400 kg of depleted uranium. These operations required over 35,000 fissile material transfers between zones and over 6000 transfers between containers. Throughout all of these movements, no mass limit violations occurred. Numerous lessons were learned over the ten year operating history. From a criticality safety perspective, the most important lesson learned was the involvement of a criticality safety practitioner in daily operations. A criticality safety engineer was assigned directly to facility operations, and was responsible for implementation of limits and controls including upkeep of the associated computerized tracking files. The criticality safety engineer was also responsible for conducting fuel handler training activities including serving on fuel handler qualification oral boards, and continually assessing operations from a criticality control perspective. The criticality safety engineer also attended bimonthly project planning meetings to identify upcoming process changes that would require criticality safety evaluation. Finally, the excellent criticality safety record was due in no small part to the continual support, involvement, trust, and confidence of project and operations mana

Chad Pope

2007-05-01T23:59:59.000Z

184

MANAGING SPENT NUCLEAR FUEL WASTES AT THE IDAHO NATIONAL LABORATORY  

SciTech Connect

The Idaho National Engineering Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy is in part due to the history of the INL as the National Reactor Testing Station, in part to its mission to recover highly enriched uranium from SNF and in part to it’s mission to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facility, some dating back 50 years in the site history. The success of the INL SNF program is measured by its ability to: 1) achieve safe existing storage, 2) continue to receive SNF from other locations, both foreign and domestic, 3) repackage SNF from wet storage to interim dry storage, and 4) prepare the SNF for dispositioning in a federal repository. Because of the diversity in the SNF and the facilities at the INL, the INL is addressing almost very condition that may exist in the SNF world. Many of solutions developed by the INL are applicable to other SNF storage sites as they develop their management strategy. The SNF being managed by the INL are in a variety of conditions, from intact assemblies to individual rods or plates to powders, rubble, and metallurgical mounts. Some of the fuel has been in wet storage for over forty years. The fuel is stored bare, or in metal cans and either wet under water or dry in vaults, caissons or casks. Inspections have shown varying degrees of corrosion and degradation of the fuel and the storage cans. Some of the fuel has been recanned under water, and the conditions of the fuel inside the second or third can are unknown. The fuel has been stored in one of 10 different facilities: five wet pools and one casks storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The wet pools range from forty years old to the most modern pool in the US Department of Energy (DOE) complex. The near-term objective is moving the fuel in the older wet storage facilities to interim dry storage facilities, thus permitting the shutdown and decommission of the older facilities. Two wet pool facilities, one at the Idaho Nuclear Technology and Engineering Center and the other at Test Area North, were targeted for initial SNF movements since these were some of the oldest at the INL. Because of the difference in the SNF materials different types of drying processes had to be developed. Passive drying, as is done with typical commercial SNF was not an option because on the condition of some of the fuel, the materials to be dried, and the low heat generation of some of the SNF. There were also size limitations in the existing facility. Active dry stations were designed to address the specific needs of the SNF and the facilities.

Hill, Thomas J

2005-09-01T23:59:59.000Z

185

EIS-0306: Treatment and Management of Sodium-Bonded Spent Nuclear Fuel |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

306: Treatment and Management of Sodium-Bonded Spent Nuclear 306: Treatment and Management of Sodium-Bonded Spent Nuclear Fuel EIS-0306: Treatment and Management of Sodium-Bonded Spent Nuclear Fuel Summary This EIS evaluates the potential environmental impacts of the proposed electrometallurgical treatment of DOE-owned sodium bonded spent nuclear fuel in the Fuel Conditioning Facility at Argonne National Laboratory-West (ANL-W). Public Comment Opportunities None available at this time. Documents Available for Download September 19, 2000 EIS-0306: Record of Decision Treatment and Management of Sodium-Bonded Spent Nuclear Fuel July 1, 2000 EIS-0306: Final Environmental Impact Statement Treatment and Management of Sodium-Bonded Spent Nuclear Fuel July 1, 1999 EIS-0306: Draft Environmental Impact Statement Treatment of Sodium-Bonded Spent Nuclear Fuel

186

Distribution of characteristics of LWR [light water reactor] spent fuel  

SciTech Connect

The purpose of this report is to develop a collective description of the entire spent fuel inventory in terms of various fuel properties relevant to Approved Testing Materials (ATMs) using information available from the Characteristics Data Base (CBD), which is sponsored by the US Department of Energy`s (DOE`s) Office of Civilian Radioactive Waste Management. A number of light-water reactor (LWR) characteristics were analyzed including assembly class representation, fuel burnup, enrichment, fuel fabrication data, defective fuel quantities, and, at PNL`s specific request, linear heat generation rate (LHGR) and the utilization of burnable poisons. A quantitative relationships was developed between burnup and enrichment for BWRs and PWRs. The relationship shows that the existing BWR ATM is near the center of the burnup-enrichment distribution, while the four PWR ATMs bracket the center of the burnup range but are on the low side of the enrichment range. Fuel fabrication data are based on vendor specifications for new fuel. Defective fuel distributions were analyzed in terms of assembly class and vendor design. LHGR values were calculated from utility data on burnup and effective full-power days; these calculations incorporate some unavoidable assumptions which may compromise the value of the results. Only a limited amount of data are available on burnable poisons at this time. Based on this distribution study, suggestions for additional ATMs are made. These are based on the class and design concepts and include BWR/2,3 barrier fuel, and the WE 17 {times} 17 class with integral burnable poison. Both should be at relatively high burnups. 16 refs., 5 figs., 15 tabs.

Reich, W.J.; Notz, K.J. [Oak Ridge National Lab., TN (USA); Moore, R.S. [Automated Sciences Group, Inc., Oak Ridge, TN (USA)

1991-01-01T23:59:59.000Z

187

Microscopic Examination of a Corrosion Front in Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

Spent uranium oxide nuclear fuel hosts a variety of trace chemical constituents, many of which must be sequestered from the biosphere during fuel storage and disposal. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within spent uranium oxide nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the plutonium and neptunium in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO{sub 2} matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl U(VI)O{sub 2}{sup 2+} mineral assemblage that is depleted in plutonium and neptunium relative to the parent fuel. At the corrosion front interface between intact fuel and the uranyl-mineral corrosion layer, we find evidence of a thin ({approx}20 micrometer) layer that is enriched in plutonium and neptunium within a predominantly U{sup 4+} environment. Available data for the standard reduction potentials for NpO{sup 2+}/Np{sup 4+} and UO{sub 2}{sup 2+}/U{sup 4+} couples indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potentials of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions. Neptunium is an important radionuclide in dose contribution according to performance assessment models of the proposed U. S. repository at Yucca Mountain, Nevada. A scientific understanding of how the UO{sub 2} matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of neptunium is needed to predict its behavior at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel [1]. In the immediate vicinity of the spent fuel's surface the redox and nucleation behavior is likely to promote/enhance nucleation of NpO{sub 2} and Np{sub 2}O{sub 5}. Alternatively, Np may be incorporated into uranyl (UO{sub 2}{sup 2+}) alteration phases [2]. In some cases, less-soluble elements such as plutonium will be enriched near the surface of the corroding fuel [3]. We have used focused synchrotron x-rays from the MRCAT beam line at the Advanced Photon Source (APS) at Argonne National Lab to examine a specimen of spent nuclear fuel that had been subject to 10 years of corrosion testing in an environment of humid air and dripping groundwater at 90 C [4]. We find evidence of a region, approximately 20 microns in thickness, enriched in plutonium and neptunium at the corrosion front that exists between the uranyl silicate alteration mineral rind and the unaltered uranium oxide fuel (Figures 1 and 2). The uranyl silicate is itself found to be depleted in these transuranic elements relative to their abundance relative to uranium in the parent fuel. This suggests a low mobility of these components owing to a resistance to oxidize further in the presence of a UO{sub 2}{sup 2+}/U{sup 4+} couple [5].

J.A> Fortner; A.J. Kropf; R.J. Finch; J.C. Cunnane

2006-06-20T23:59:59.000Z

188

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network (OSTI)

neutrino Production at Nuclear Reactors Z. Djurcic 1 , ?emission rates from nuclear reactors are determined fromlarge commercial nuclear reactors are playing an important

Djurcic, Zelimir

2009-01-01T23:59:59.000Z

189

RADIATION FACILITY FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

A radiation facility is designed for irradiating samples in close proximity to the core of a nuclear reactor. The facility comprises essentially a tubular member extending through the biological shield of the reactor and containing a manipulatable rod having the sample carrier at its inner end, the carrier being longitudinally movable from a position in close proximity to the reactor core to a position between the inner and outer faces of the shield. Shield plugs are provided within the tubular member to prevent direct radiation from the core emanating therethrough. In this device, samples may be inserted or removed during normal operation of the reactor without exposing personnel to direct radiation from the reactor core. A storage chamber is also provided within the radiation facility to contain an irradiated sample during the period of time required to reduce the radioactivity enough to permit removal of the sample for external handling. (AEC)

Currier, E.L. Jr.; Nicklas, J.H.

1961-12-12T23:59:59.000Z

190

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

SciTech Connect

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

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

1996-10-01T23:59:59.000Z

191

Integral Fast Reactor: A future source of nuclear energy  

SciTech Connect

Argonne National Laboratory is developing a reactor concept that would be an important part of the worlds energy future. This report discusses the Integral Fast Reactor (IFR) concept which provides significant improvements over current generation reactors in reactor safety, plant complexity, nuclear proliferation, and waste generation. Two major facilities, a reactor and a fuel cycle facility, make up the IFR concept. The reactor uses fast neutrons and metal fuel in a sodium coolant at atmospheric pressure that relies on laws of physics to keep it safe. The fuel cycle facility is a hot cell using remote handling techniques for fabricating reactor fuel. The fuel feed stock includes spent fuel from the reactor, and potentially, spent light water reactor fuel and plutonium from weapons. This paper discusses the unique features of the IFR concept and the differences the quality assurance program has from current commercial practices. The IFR concept provides an opportunity to design a quality assurance program that makes use of the best contemporary ideas on management and quality.

Southon, R.

1993-09-01T23:59:59.000Z

192

NUCLEAR REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

Currier, E.L. Jr.; Nicklas, J.H.

1963-06-11T23:59:59.000Z

193

EM Safely and Efficiently Manages Spent Nuclear Fuel | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Services » Waste Management » Nuclear Materials & Waste » EM Services » Waste Management » Nuclear Materials & Waste » EM Safely and Efficiently Manages Spent Nuclear Fuel EM Safely and Efficiently Manages Spent Nuclear Fuel Dry storage casks at Idaho National Laboratory can safely house spent nuclear fuel for decades. Dry storage casks at Idaho National Laboratory can safely house spent nuclear fuel for decades. EM's mission is to safely and efficiently manage its spent nuclear fuel and prepare it for disposal in a geologic repository. Previously, the Office of Environmental Management's (EM) mission had included the safe and efficient management of its spent nuclear fuel (SNF) and preparation for its disposal in a geologic repository. However, in May 2009, the planned geologic repository at Yucca Mountain was cancelled. The

194

Technical strategy for the management of INEEL spent nuclear fuel  

SciTech Connect

This report presents evaluations, findings, and recommendations of the Idaho National Engineering and Environmental Laboratory (INEEL) Spent Nuclear Fuel Task Team. The technical strategy developed by the Task Team includes stabilization, near term storage, packaging, transport, and ultimate disposal. Key issues identified and discussed include waste characterization, criticality, packaging, waste form performance, and special fuels. Current plans focus on onsite needs, and include three central elements: (1) resolution of near-term vulnerabilities, (2) consolidation of storage locations, and (3) achieving dry storage in transportable packages. In addition to the Task Team report, appendices contain information on the INEEL spent fuel inventory; regulatory decisions and agreements; and analyses of criticality, packaging, storage, transportation, and system performance of a geological repository. 16 refs., 6 figs., 4 tabs.

1997-03-01T23:59:59.000Z

195

HOMOGENEOUS NUCLEAR REACTOR  

SciTech Connect

This homogeneous reactor comprises a core occupied by a solution of a fissile material in a moderator liquid and a breeder region enclosing the core and having a suspension of fertile material in the same moderator liquid. There is communication between the core and breeder to allow mass transfer and pressure equalization between the regions. The zones each have a separate circuit for removing heat by a mixer chamber situated inside the reactor vessel. The effluents coming from the two regions are mixed and led to a common device for separation into a clear solution and suspension, which are each led back to its corresponding circuit. To control the relative concentration of the two regions, an evaporator is provided separating a part of the moderator liquid from the solution occupying the core, the condensed separated moderator liquid being led into the breeder region. (NPO)

1960-07-11T23:59:59.000Z

196

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

SciTech Connect

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

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

1992-04-01T23:59:59.000Z

197

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

SciTech Connect

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

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

1992-04-01T23:59:59.000Z

198

Nuclear Reactors and Technology; (USA)  

SciTech Connect

Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

Cason, D.L.; Hicks, S.C. (eds.)

1991-01-01T23:59:59.000Z

199

METHOD OF OPERATING NUCLEAR REACTORS  

DOE Patents (OSTI)

A method is presented for obtaining enhanced utilization of natural uranium in heavy water moderated nuclear reactors by charging the reactor with an equal number of fuel elements formed of natural uranium and of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction. The reactor is operated until the rate of burnup of plutonium equals its rate of production, the fuel elements are processed to recover plutonium, the depleted uranium is discarded, and the remaining uranium is formed into fuel elements. These fuel elements are charged into a reactor along with an equal number of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction, and reuse of the uranium is continued as aforesaid until it wlll no longer support a chain reaction when combined with an equal quantity of natural uranium.

Untermyer, S.

1958-10-14T23:59:59.000Z

200

Advanced Nuclear Reactors | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Advanced Nuclear Advanced Nuclear Reactors Advanced Nuclear Reactors Turbulent Flow of Coolant in an Advanced Nuclear Reactor Visualizing Coolant Flow in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor vessel where all the nuclear reactions occur to produce the heat and steam used for power generation. Nonetheless, an entire power plant consists of many other support components and key structures like coolant pipes; pumps and tanks including their surrounding steel framing; and concrete containment and support structures. The Reactors Product Line within NEAMS is concerned with modeling the reactor vessel as well as those components of a complete power plant that

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Closure Mechanism and Method for Spent Nuclear Fuel Canisters  

DOE Patents (OSTI)

A canister is provided for storing, transporting, and/or disposing of spent nuclear fuel. The canister includes a canister shell, a top shield plug disposed within the canister, and a leak-tight closure arrangement. The closure arrangement includes a shear ring which forms a containment boundary of the canister, and which is welded to the canister shell and top shield plug. An outer seal plate, forming an outer seal, is disposed above the shear ring and is welded to the shield plug and the canister.

Doman, Marvin J.

2004-11-23T23:59:59.000Z

202

Method for reprocessing and separating spent nuclear fuels  

DOE Patents (OSTI)

Spent nuclear fuels, including actinide fuels, volatile and non-volatile 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.

Krikorian, Oscar H. (Danville, CA); Grens, John Z. (Livermore, CA); Parrish, Sr., William H. (Walnut Creek, CA)

1983-01-01T23:59:59.000Z

203

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

DOE Patents (OSTI)

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.

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

1982-01-19T23:59:59.000Z

204

Closure mechanism and method for spent nuclear fuel canisters  

DOE Patents (OSTI)

A canister is provided for storing, transporting, and/or disposing of spent nuclear fuel. The canister includes a canister shell, a top shield plug disposed within the canister, and a leak-tight closure arrangement. The closure arrangement includes a shear ring which forms a containment boundary of the canister, and which is welded to the canister shell and top shield plug. An outer seal plate, forming an outer seal, is disposed above the shear ring and is welded to the shield plug and the canister.

Doman, Marvin J. (Monroeville, PA)

2004-11-23T23:59:59.000Z

205

Corrosion of Spent Nuclear Fuel: The Long-Term Assessment  

Science Conference Proceedings (OSTI)

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

Rodney C. Ewing

2004-10-07T23:59:59.000Z

206

Economic analysis of nuclear reactors  

SciTech Connect

The report presents several methods for estimating the power costs of nuclear reactors. When based on a consistent set of economic assumptions, total power costs may be useful in comparing reactor alternatives. The principal items contributing to the total power costs of a nuclear power plant are: (1) capital costs, (2) fuel cycle costs, (3) operation and maintenance costs, and (4) income taxes and fixed charges. There is a large variation in capital costs and fuel expenses among different reactor types. For example, the standard once-through LWR has relatively low capital costs; however, the fuel costs may be very high if U/sub 3/O/sub 8/ is expensive. In contrast, the FBR has relatively high capital costs but low fuel expenses. Thus, the distribution of expenses varies significantly between these two reactors. In order to compare power costs, expenses and revenues associated with each reactor may be spread over the lifetime of the plant. A single annual cost, often called a levelized cost, may be obtained by the methods described. Levelized power costs may then be used as a basis for economic comparisons. The paper discusses each of the power cost components. An exact expression for total levelized power costs is derived. Approximate techniques of estimating power costs will be presented.

Owen, P.S.; Parker, M.B.; Omberg, R.P.

1979-05-01T23:59:59.000Z

207

The burnup dependence of light water reactor spent fuel oxidation  

SciTech Connect

Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5).

Hanson, B.D.

1998-07-01T23:59:59.000Z

208

Guidelines for Fabrication, Examination, Testing and Oversight of Spent Nuclear Fuel Dry Storage Systems  

Science Conference Proceedings (OSTI)

The Nuclear Waste Policy Act (NWPA) of 1982 and subsequent amendments require the U. S. Department of Energy (DOE) to receive and be responsible for disposal of spent commercial nuclear power plant fuel from U.S. utilities. However, because of delays in the siting of a permanent federal repository, and with no federal interim storage facilities designated, U.S. utilities have been forced to provide additional spent nuclear fuel (SNF) storage capability to accommodate spent fuel discharge requirements. At...

1999-12-10T23:59:59.000Z

209

Reactor and Nuclear Systems Division (RNSD)  

NLE Websites -- All DOE Office Websites (Extended Search)

RNSD Home RNSD Home Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Staff Details (CV/Bios) Publications Org Chart Contact Us ORNL Staff Only Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Reactor and Nuclear Systems Division News Highlights U.S. Rep. Fleischmann touts ORNL as national energy treasure Martin Peng wins Fusion Power Associates Leadership Award

210

U.S. Spent Nuclear Fuel Data as of December 31, 2002  

U.S. Energy Information Administration (EIA)

Spent Nuclear Fuel: The Energy Information Administration (EIA) collects data for the Office of Civilian Radioactive Waste Management (OCRWM) on the detailed ...

211

Instrumented, Shielded Test Canister System for Evaluation of Spent Nuclear Fuel in Dry Storage  

SciTech Connect

This document describes the development of an instrumented, shielded test canister system to store and monitor aluminum-based spent nuclear duel under dry storage conditions.

Sindelar, R.L.

1999-10-21T23:59:59.000Z

212

Behavior of Spent Nuclear Fuel in Water Pool Storage  

Office of Scientific and Technical Information (OSTI)

Behavior of Spent Nuclear Behavior of Spent Nuclear Fuel in Water Pool Storage A. 0; Johnson, jr. , I ..: . Prepared Cor the Energy Research and Development Administration under Contract EY-76-C-06-1830 ---- Pat t i ~ < N ~ ~ r ~ t b w t ~ - ! I , ~ I ~ ~ ~ I . I I ~ ) ~ I I ~ ~ N O T I C E T€& - was prepad pnpn4. m w n t of w k spon-d by the Unitd S t . & ) C a u n m ~ (*WU ij*. M t e d $tam w the Wqy R e s e w & a d Ohrsropmcnt ~dmhirmlion, nor m y d thair ewhew,,nq Pny @fw a n t r ~ ~ t 0 ~ 1 , s ~ k m r i t r i l t t q r , ~ , m r tWf ernpfQw, r(tLltm any wartany, s x p r e s or kWld,= w w aAql -9 . o r r w p a m l ~ ~ t y for e~ o r uodruincvr of any infomutim, 9 F p d + d - , or repratants that -would nat 1 d - e privately owned rfghas. ,i PAQFIC NORTHWEST UBORATORY operated b ;"' SArnLLE ' fw the E M R m RESEARCH AND DEVELOPMENT ADMINISTRAT1QN Wk.Cwfraa rv-76c-ts-is38

213

Propellant actuated nuclear reactor steam depressurization valve  

DOE Patents (OSTI)

A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

Ehrke, Alan C. (San Jose, CA); Knepp, John B. (San Jose, CA); Skoda, George I. (Santa Clara, CA)

1992-01-01T23:59:59.000Z

214

Reactors for nuclear electric propulsion  

SciTech Connect

Propulsion is the key to space exploitation and power is the key to propulsion. This paper examines the role of nuclear fission reactors as the primary power source for high specific impulse electric propulsion systems for space missions of the 1980s and 1990s. Particular mission applications include transfer to and a reusable orbital transfer vehicle from low-Earth orbit to geosynchronous orbit, outer planet exploration and reconnaissance missions, and as a versatile space tug supporting lunar resource development. Nuclear electric propulsion is examined as an indispensable component in space activities of the next two decades.

Buden, D.; Angelo, J.A. Jr.

1981-01-01T23:59:59.000Z

215

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

Bassett, C.H.

1961-05-16T23:59:59.000Z

216

Flow duct for nuclear reactors  

DOE Patents (OSTI)

Improved liquid sodium flow ducts for nuclear reactors are described wherein the improvement comprises varying the wall thickness of each of the walls of a polygonal tubular duct structure so that each of the walls is of reduced cross-section along the longitudinal center line and of a greater cross-section along wall junctions with the other walls to form the polygonal tubular configuration.

Straalsund, Jerry L. (Richland, WA)

1978-01-01T23:59:59.000Z

217

NUCLEAR REACTOR COMPENENT CLADDING MATERIAL  

DOE Patents (OSTI)

Fuel elements and coolant tubes used in nuclear reactors of the heterogeneous, water-cooled type are described, wherein the coolant tubes extend through the moderator and are adapted to contain the fuel elements. The invention comprises forming the coolant tubes and the fuel element cladding material from an alloy of aluminum and nickel, or an alloy of aluminum, nickel, alloys are selected to prevent intergranular corrosion of these components by water at temperatures up to 35O deg C.

Draley, J.E.; Ruther, W.E.

1959-01-27T23:59:59.000Z

218

A Technical Review of Non-Destructive Assay Research for the Characterization of Spent Nuclear Fuel Assemblies Being Conducted Under the US DOE NGSI - 11544  

E-Print Network (OSTI)

the Characterization of Spent Nuclear Fuel Assemblies BeingSociety’s Advances in Nuclear Fuel Management IV, HiltonPlutonium Mass in Spent Nuclear Fuel,” 2010 ANS Annual

Croft, S.

2012-01-01T23:59:59.000Z

219

Thermal analysis for a spent reactor fuel storage test in granite  

Science Conference Proceedings (OSTI)

A test is conducted in which spent fuel assemblies from an operating commercial nuclear power reactor are emplaced in the Climax granite at the US Department of Energy`s Nevada Test Site. In this generic test, 11 canisters of spent PWR fuel are emplaced vertically along with 6 electrical simulator canisters on 3 m centers, 4 m below the floor of a storage drift which is 420 m below the surface. Two adjacent parallel drifts contain electrical heaters, operated to simulate (in the vicinity of the storage drift) the temperature fields of a large repository. This test, planned for up to five years duration, uses fairly young fuel (2.5 years out of core) so that the thermal peak will occur during the time frame of the test and will not exceed the peak that would not occur until about 40 years of storage had older fuel (5 to 15 years out of core) been used. This paper describes the calculational techniques and summarizes the results of a large number of thermal calculations used in the concept, basic design and final design of the spent fuel test. The results of the preliminary calculations show the effects of spacing and spent fuel age. Either radiation or convection is sufficient to make the drifts much better thermal conductors than the rock that was removed to create them. The combination of radiation and convection causes the drift surfaces to be nearly isothermal even though the heat source is below the floor. With a nominal ventilation rate of 2 m{sup 3}/s and an ambient rock temperature of 23{sup 0}C, the maximum calculated rock temperature (near the center of the heat source) is about 100{sup 0}C while the maximum air temperature in the drift is around 40{sup 0}C. This ventilation (1 m{sup 3}/s through the main drift and 1/2 m{sup 3}/s through each of the side drifts) will remove about 1/3 of the heat generated during the first five years of storage.

Montan, D.N.

1980-09-01T23:59:59.000Z

220

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

Science Conference Proceedings (OSTI)

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

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

2009-01-01T23:59:59.000Z

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221

Commercial Spent Nuclear Fuel Waste Package Misload Analysis  

Science Conference Proceedings (OSTI)

The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis Department. Before using the results of this calculation, the reader is cautioned to verify that the assumptions made in this calculation regarding the waste stream, the loading process, and the staging of the spent nuclear fuel assemblies are applicable.

A. Alsaed

2005-07-28T23:59:59.000Z

222

DOE-owned spent nuclear fuel program plan  

SciTech Connect

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

1995-11-01T23:59:59.000Z

223

Performance assessment of DOE spent nuclear fuel and surplus plutonium  

SciTech Connect

Yucca Mountain, in southern Nevada, is under consideration by the US Department of Energy (DOE) as a potential site for the disposal of the nation`s radioactive wastes in a geologic repository. The wastes consist of commercial spent fuel, DOE spent nuclear fuel (SNF), high level waste (HLW), and surplus plutonium. The DOE was mandated by Congress in the fiscal 1997 Energy and Water Appropriations Act to complete a viability assessment (VA) of the repository in September of 1998. The assessment consists of a preliminary design concept for the critical elements of the repository, a total system performance assessment (TSPA), a plan and cost estimate for completion of the license application, and an estimate of the cost to construct and operate the repository. This paper presents the results of the sensitivity analyses that were conducted to examine the behavior of DOE SNF and plutonium waste forms in the environment of the base case repository that was modeled for the TSPA-VA. Fifteen categories of DOE SNF and two Plutonium waste forms were examined and their contribution to radiation dose to humans was evaluated.

Duguid, J.O.; Vallikat, V.; McNeish, J.

1998-01-01T23:59:59.000Z

224

Energy Secretary to Visit Georgia Nuclear Reactor Site and Tennessee...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Nuclear Reactor Site and Tennessee Laboratory to Highlight Administration Support for Nuclear Energy Energy Secretary to Visit Georgia Nuclear Reactor Site and Tennessee...

225

Spent Nuclear Fuel Dry Transfer System Cold Demonstration Project Final Report  

SciTech Connect

The spent nuclear fuel dry transfer system (DTS) provides an interface between large and small casks and between storage-only and transportation casks. It permits decommissioning of reactor pools after shutdown and allows the use of large storage-only casks for temporary onsite storage of spent nuclear fuel irrespective of reactor or fuel handling limitations at a reactor site. A cold demonstration of the DTS prototype was initiated in August 1996 at the Idaho National Engineering and Environmental Laboratory (INEEL). The major components demonstrated included the fuel assembly handling subsystem, the shield plug/lid handling subsystem, the cask interface subsystem, the demonstration control subsystem, a support frame, and a closed circuit television and lighting system. The demonstration included a complete series of DTS operations from source cask receipt and opening through fuel transfer and closure of the receiving cask. The demonstration included both normal operations and recovery from off-normal events. It was designed to challenge the system to determine whether there were any activities that could be made to jeopardize the activities of another function or its safety. All known interlocks were challenged. The equipment ran smoothly and functioned as designed. A few "bugs" were corrected. Prior to completion of the demonstration testing, a number of DTS prototype systems were modified to apply lessons learned to date. Additional testing was performed to validate the modifications. In general, all the equipment worked exceptionally well. The demonstration also helped confirm cost estimates that had been made at several points in the development of the system.

Christensen, Max R; McKinnon, M. A.

1999-12-01T23:59:59.000Z

226

Reactivity control assembly for nuclear reactor  

DOE Patents (OSTI)

Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

Bollinger, Lawrence R. (Schenectady, NY)

1984-01-01T23:59:59.000Z

227

Eddy Current Examination of Spent Nuclear Fuel Canister Closure Welds  

SciTech Connect

The National Spent Nuclear Fuel Program (NSNFP) has developed standardized DOE SNF canisters for handling and interim storage of SNF at various DOE sites as well as SNF transport to and SNF handling and disposal at the repository. The final closure weld of the canister will be produced remotely in a hot cell after loading and must meet American Society of Mechanical Engineers (ASME) Section III, Division 3 code requirements thereby requiring volumetric and surface nondestructive evaluation to verify integrity. This paper discusses the use of eddy current testing (ET) to perform surface examination of the completed welds and repair cavities. Descriptions of integrated remote welding/inspection system and how the equipment is intended function will also be discussed.

Arthur D. Watkins; Dennis C. Kunerth; Timothy R. McJunkin

2006-04-01T23:59:59.000Z

228

Electric Power Produced from Nuclear Reactor | National Nuclear Security  

NLE Websites -- All DOE Office Websites (Extended Search)

Electric Power Produced from Nuclear Reactor | National Nuclear Security Electric Power Produced from Nuclear Reactor | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > Electric Power Produced from Nuclear Reactor Electric Power Produced from Nuclear Reactor December 20, 1951 Arco, ID Electric Power Produced from Nuclear Reactor

229

Burnup verification measurements on spent fuel assemblies at Arkansas Nuclear One  

SciTech Connect

Burnup verification measurements have been performed using the Fork system at Arkansas Nuclear One, Units 1 and 2, operated by Energy Operations, Inc. Passive neutron and gamma-ray measurements on individual spent fuel assemblies were correlated with the reactor records for burnup, cooling time, and initial enrichment. The correlation generates an internal calibration for the system in the form of a power law determined by a least squares fit to the neutron data. The values of the exponent in the power laws were 3.83 and 4.35 for Units 1 and 2, respectively. The average deviation of the reactor burnup records from the calibration determined from the measurements is a measure of the random error in the burnup records. The observed average deviations were 2.7% and 3.5% for assemblies at Units 1 and 2, respectively, indicating a high degree of consistency in the reactor records. Two non-standard assemblies containing neutron sources were studied at Unit 2. No anomalous measurements were observed among the standard assemblies at either Unit. The effectiveness of the Fork system for verification of reactor records is due to the sensitivity of the neutron yield to burnup, the self-calibration generated by a series of measurements, the redundancy provided by three independent detection systems, and the operational simplicity and flexibility of the design.

Ewing, R.I.

1995-09-01T23:59:59.000Z

230

Fusion option to dispose of spent nuclear fuel and transuranic elements  

SciTech Connect

The fusion option is examined to solve the disposition problems of the spent nuclear fuel and the transuranic elements. The analysis of this report shows that the top rated solution, the elimination of the transuranic elements and the long-lived fission products, can be achieved in a fusion reactor. A 167 MW of fusion power from a D-T plasma for sixty years with an availability factor of 0.75 can transmute all the transuranic elements and the long-lived fission products of the 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. The operating time can be reduced to thirty years with use of 334 MW of fusion power, a system study is needed to define the optimum time. In addition, the fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future. Fusion blankets with a liquid carrier for the transuranic elements can achieve a transmutation rate for the transuranic elements up to 80 kg/MW.y of fusion power with k{sub eff} of 0.98. In addition, the liquid blankets have several advantages relative to the other blanket options. The energy from this transmutation is utilized to produce revenue for the system. Molten salt (Flibe) and lithium-lead eutectic are identified as the most promising liquids for this application, both materials are under development for future fusion blanket concepts. The Flibe molten salt with transuranic elements was developed and used successfully as nuclear fuel for the molten salt breeder reactor in the 1960's.

Gohar, Y.

2000-02-10T23:59:59.000Z

231

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

A nuclear reactor fuel element comprising high density ceramic fissionable material enclosed in a tubular cladding of corrosion-resistant material is described. The fissionable material is in the form of segments of a tube which have cooperating tapered interfaces which produce outward radial displacement when the segments are urged axially together. A resilient means is provided within the tubular housing to constantly urge the fuel segments axially. This design maintains the fuel material in tight contacting engagement against the inner surface of the outer cladding tube to eliminate any gap therebetween which may be caused by differential thermal expansion between the fuel material and the material of the tube.

Bassett, C.H.

1961-05-01T23:59:59.000Z

232

Fate of Noble Metals during the Pyroprocessing of Spent Nuclear Fuel  

SciTech Connect

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

B.R. Westphal; D. Vaden; S.X. Li; G.L. Fredrickson; R.D. Mariani

2009-09-01T23:59:59.000Z

233

Sampling and analysis plan for the preoperational environmental survey of the spent nuclear fuel project facilities  

Science Conference Proceedings (OSTI)

This sampling and analysis plan will support the preoperational environmental monitoring for construction, development, and operation of the Spent Nuclear Fuel (SNF) Project facilities, which have been designed for the conditioning and storage of spent nuclear fuels; particularly the fuel elements associated with the operation of N-Reactor. The SNF consists principally of irradiated metallic uranium, and therefore includes plutonium and mixed fission products. The primary effort will consist of removing the SNF from the storage basins in K East and K West Areas, placing in multicanister overpacks, vacuum drying, conditioning, and subsequent dry vault storage in the 200 East Area. The primary purpose and need for this action is to reduce the risks to public health and safety and to the environment. Specifically these include prevention of the release of radioactive materials into the air or to the soil surrounding the K Basins, prevention of the potential migration of radionuclides through the soil column to the nearby Columbia River, reduction of occupational radiation exposure, and elimination of the risks to the public and to workers from the deterioration of SNF in the K Basins.

MITCHELL, R.M.

1999-04-01T23:59:59.000Z

234

Reactors: Modern-Day Alchemy - Argonne's Nuclear Science and Technology  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements > Achievements > Legacy > Reactors: Modern-Day Alchemy About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

235

Achievements: Nuclear Reactors designed/built by Argonne National  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements > Achievements > Argonne National Laboratory Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

236

The Czech National R&D Program of Nuclear Incineration of PWR Spent Fuel in a Transmuter with Liquid Fuel  

E-Print Network (OSTI)

The principle drawbacks of any kind of solid nuclear fuel are listed and briefly analysed in the first part of the paper. On the basis of this analysis, the liquid fuel concept and its benefits are introduced and briefly described in the following parts of the paper allowing to develop new reactor systems for nuclear incineration of spent fuel from conventional reactors and a new clean source of energy. As one of the first realistic attempts to utilise the advantages of liquid fuel, the reactor/blanket system with molten fluoride salts in the role of fuel and coolant simultaneously, as incorporated in the accelerator-driven transmutation technology (ADTT) being proposed in [1], has been proposed for a deeper, both theoretical and experimental studies in [2]. There will be a preliminary design concept of an experimental assembly LA-0 briefly introduced in the paper which is under preparation in the Czech Republic for such a project [3]. 1

M. Hron

1998-01-01T23:59:59.000Z

237

Simulated nuclear reactor fuel assembly  

DOE Patents (OSTI)

An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

Berta, Victor T. (Idaho Falls, ID)

1993-01-01T23:59:59.000Z

238

Spent nuclear fuels project characterization data quality objectives strategy  

SciTech Connect

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

Lawrence, L.A.; Thornton, T.A. [Pacific Northwest Lab., Richland, WA (United States); Redus, K.S.

1994-12-01T23:59:59.000Z

239

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

Not Available

1991-04-01T23:59:59.000Z

240

Nuclear reactor characteristics and operational history  

U.S. Energy Information Administration (EIA) Indexed Site

Nuclear > U.S. reactor operation status tables Nuclear > U.S. reactor operation status tables Nuclear Reactor Operational Status Tables Release date: November 22, 2011 Next release date: November 2012 See also: Table 2. Ownership Data, Table 3. Characteristics and Operational History Table 1. Nuclear Reactor, State, Type, Net Capacity, Generation, and Capacity Factor PDF XLS Plant/Reactor Name Generator ID State Type 2009 Summer Capacity Net MW(e)1 2010 Annual Generation Net MWh2 Capacity Factor Percent3 Arkansas Nuclear One 1 AR PWR 842 6,607,090 90 Arkansas Nuclear One 2 AR PWR 993 8,415,588 97 Beaver Valley 1 PA PWR 892 7,119,413 91 Beaver Valley 2 PA PWR 885 7,874,151 102 Braidwood Generation Station 1 IL PWR 1,178 9,196,689 89

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Nuclear reactor characteristics and operational history  

Gasoline and Diesel Fuel Update (EIA)

Nuclear > U.S. reactor operation status tables Nuclear > U.S. reactor operation status tables Nuclear Reactor Operational Status Tables Release date: November 22, 2011 Next release date: November 2012 See also: Table 1. Capacity and Generation, Table 2. Ownership Data Table 3. Nuclear Reactor Characteristics and Operational History PDF XLS Plant Name Generator ID Type Reactor Supplier and Model Construction Start Grid Connection Original Expiration Date License Renewal Application License Renewal Issued Extended Expiration Arkansas Nuclear One 1 PWR Babcock&Wilcox, Lower Loop 10/1/1968 8/17/1974 5/20/2014 2/1/2000 6/20/2001 5/20/2034 Arkansas Nuclear One 2 PWR Combustion Eng. 7/1/1971 12/26/1978 7/17/2018 10/15/2003 6/30/2005 7/17/2038

242

HOW MANY DID YOU SAY? HISTORICAL AND PROJECTED SPENT NUCLEAR FUEL SHIPMENTS IN THE UNITED STATES, 1964 - 2048  

Science Conference Proceedings (OSTI)

No comprehensive, up-to-date, official database exists for spent nuclear fuel shipments in the United States. The authors review the available data sources, and conclude that the absence of such a database can only be rectified by a major research effort, similar to that carried out by Oak Ridge National Laboratory (ORNL) in the early 1990s. Based on a variety of published references, and unpublished data from the U.S. Nuclear Regulatory Commission (NRC), the authors estimate cumulative U.S. shipments of commercial spent fuel for the period 1964-2001. The cumulative estimates include quantity shipped, number of cask-shipments, and shipment-miles, by truck and by rail. The authors review previous estimates of future spent fuel shipments, including contractor reports prepared for the U.S. Department of Energy (DOE), NRC, and the State of Nevada. The DOE Final Environmental Impact Statement (FEIS) for Yucca Mountain includes projections of spent nuclear fuel and high-level radioactive was te shipments for two inventory disposal scenarios (24 years and 38 years) and two national transportation modal scenarios (''mostly legal-weight truck'' and ''mostly rail''). Commercial spent fuel would compromise about 90 percent of the wastes shipped to the repository. The authors estimate potential shipments to Yucca Mountain over 38 years (2010-2048) for the DOE ''mostly legal-weight truck'' and ''mostly rail'' scenarios, and for an alternative modal mix scenario based on current shipping capabilities of the 72 commercial reactor sites. The cumulative estimates of future spent fuel shipments include quantity shipped, number of cask-shipments, and shipment-miles, by legal-weight truck, heavy-haul truck, rail and barge.

Halstead, Robert J.; Dilger, Fred

2003-02-27T23:59:59.000Z

243

EIA - State Nuclear Profiles - Energy Information Administration  

U.S. Energy Information Administration (EIA)

Uranium fuel, nuclear reactors, generation, spent fuel. Total Energy. Comprehensive data summaries, comparisons, analysis, and projections integrated across all ...

244

Nuclear Regulatory Commission approves construction of ...  

U.S. Energy Information Administration (EIA)

Uranium fuel, nuclear reactors, generation, spent fuel. Total Energy. ... near Spring City, Tennessee after initial construction began in 1973. ...

245

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

SciTech Connect

On June 1, 1995, DOE issued a Record of Decision [60 Federal Register 28680] for the Department-wide management of spent nuclear fuel (SNF); regionalized storage of SNF by fuel type was selected as the preferred alternative. The proposed action evaluated in this environmental assessment is the management of SNF on the Oak Ridge Reservation (ORR) to implement this preferred alternative of regional storage. SNF would be retrieved from storage, transferred to a hot cell if segregation by fuel type and/or repackaging is required, loaded into casks, and shipped to off-site storage. The proposed action would also include construction and operation of a dry cask SNF storage facility on ORR, in case of inadequate SNF storage. Action is needed to enable DOE to continue operation of the High Flux Isotope Reactor, which generates SNF. This report addresses environmental impacts.

1996-02-01T23:59:59.000Z

246

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

SciTech Connect

CH2M HILL Plateau Remediation Company (CHPRC) and AREVA Federal Services, LLC (AFS) have been working collaboratively to develop and deploy technologies to remove, transport, and interim store remote-handled sludge from the 10S-K West Reactor Fuel Storage Basin on the U.S. Department of Energy (DOE) Hanford Site near Richland, WA, USA. Two disposal paths exist for the different types of sludge found in the K West (KW) Basin. One path is to be managed as Spent Nuclear Fuel (SNF) with eventual disposal at an SNF at a yet to be licensed repository. The second path will be disposed as remote-handled transuranic (RH-TRU) waste at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, NM. This paper describes the systems developed and executed by the Knockout Pot (KOP) Disposition Subproject for processing and interim storage of the sludge managed as SNF, (i.e., KOP material).

Raymond, R. E. [CH2M HIll Plateau Remediation Company, Richland, WA (United States); Evans, K. M. [AREVA, Avignon (France)

2012-10-22T23:59:59.000Z

247

Results from Nevada Nuclear Waste Storage Investigations (NNWSI) Series 3 spent fuel dissolution tests  

SciTech Connect

The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Yucca Mountain Project (YMP), formerly the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Specimens prepared from pressurized water reactor fuel rod segments were tested in sealed stainless steel vessels in Nevada Test Site J-13 well water at 85{degree}C and 25{degree}C. The test matrix included three specimens of bare-fuel particles plus cladding hulls, two fuel rod segments with artificially defected cladding and water-tight end fittings, and an undefected fuel rod section with watertight end fittings. Periodic solution samples were taken during test cycles with the sample volumes replenished with fresh J-13 water. Test cycles were periodically terminated and the specimens restarted in fresh J-13 water. The specimens were run for three cycles for a total test duration of 15 months. 22 refs., 32 figs., 26 tabs.

Wilson, C.N.

1990-06-01T23:59:59.000Z

248

Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)  

Science Conference Proceedings (OSTI)

High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next Generation Safeguards Initiative (NGSI).

Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

2009-10-01T23:59:59.000Z

249

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect

This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

Not Available

1993-11-01T23:59:59.000Z

250

The Department of Energy's Spent Nuclear Fuel Canisters andTransporta...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

nuclear fuel generated from research and development, plutonium production, and the Naval Nuclear Propulsion Program (Naval Reactors). Under current national policy, the Department...

251

Nuclear Reactor Technologies | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Technologies Reactor Technologies Nuclear Reactor Technologies TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Small Modular Reactor Technologies Small modular reactors can also be made in factories and transported to sites where they would be ready to "plug and play" upon arrival, reducing both capital costs and construction times. The smaller size also makes these reactors ideal for small electric grids and for locations that

252

Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant  

SciTech Connect

The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

Guenther, R.J.; Johnson, A.B. Jr.; Lund, A.L.; Gilbert, E.R. [and others

1994-11-01T23:59:59.000Z

253

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

SciTech Connect

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

1994-06-01T23:59:59.000Z

254

SCALE Validation Experience Using an Expanded Isotopic Assay Database for Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

The availability of measured isotopic assay data to validate computer code predictions of spent fuel compositions applied in burnup-credit criticality calculations is an essential component for bias and uncertainty determination in safety and licensing analyses. In recent years, as many countries move closer to implementing or expanding the use of burnup credit in criticality safety for licensing, there has been growing interest in acquiring additional high-quality assay data. The well-known open sources of assay data are viewed as potentially limiting for validating depletion calculations for burnup credit due to the relatively small number of isotopes measured (primarily actinides with relatively few fission products), sometimes large measurement uncertainties, incomplete documentation, and the limited burnup and enrichment range of the fuel samples. Oak Ridge National Laboratory (ORNL) recently initiated an extensive isotopic validation study that includes most of the public data archived in the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) electronic database, SFCOMPO, and new datasets obtained through participation in commercial experimental programs. To date, ORNL has analyzed approximately 120 different spent fuel samples from pressurized-water reactors that span a wide enrichment and burnup range and represent a broad class of assembly designs. The validation studies, completed using SCALE 5.1, are being used to support a technical basis for expanded implementation of burnup credit for spent fuel storage facilities, and other spent fuel analyses including radiation source term, dose assessment, decay heat, and waste repository safety analyses. This paper summarizes the isotopic assay data selected for this study, presents validation results obtained with SCALE 5.1, and discusses some of the challenges and experience associated with evaluating the results. Preliminary results obtained using SCALE 6 and ENDF/B-VII cross sections libraries are also briefly summarized. Oak Ridge National Laboratory (ORNL) has been performing spent-fuel isotopic validation studies using the depletion analysis methods in the SCALE [1] code system for the past 20 years. These studies involve comparisons of calculated inventories against measured isotopic composition data obtained from destructive radiochemical analysis of commercial spent nuclear fuel samples. The results of these benchmark studies are used to quantify the bias and uncertainties associated with isotopic calculations and ultimately determine appropriate margins for uncertainty that can be applied in safety-related analyses such as burnup credit in criticality calculations, decay heat analysis, and source terms. Previous studies using several versions of SCALE and nuclear data libraries have been published in multiple validation reports [2-6] that evaluate selected experimental data obtained largely from public sources. A study was recently initiated at ORNL with the objectives of updating and expanding the validation calculations using a comprehensive database of experimental isotopic assay data that includes isotopic composition data obtained from both publicly available sources and international commercial programs. As part of the study, an extensive isotopic database of nearly 120 measured spent fuel samples with an expanded range of initial enrichments and burnup values compared to previously analyzed data was reviewed and analyzed. The calculations were performed using two-dimensional (2-D) assembly models and a consistent set of modeling assumptions using the SCALE 5.1 code system and ENDF/B-V 44-group cross section library. As part of the current study, detailed benchmark modeling information and measurement data are being documented in a format that is readily usable for validating depletion and decay codes. The work is being extended to include analysis results using SCALE 6 and the ENDF/B-VII 238-group cross section library. This paper describes the isotopic composition data evaluated in this study and highlights the pre

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

2009-01-01T23:59:59.000Z

255

Small Modular Nuclear Reactors | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Technologies » Small Modular Reactor Technologies » Small Modular Nuclear Reactors Small Modular Nuclear Reactors Cutaway of 2-Unit Generation mPower SMR Installation. | © 2012 Generation mPower LLC. All Rights Reserved. Reprinted with permission. Cutaway of 2-Unit Generation mPower SMR Installation. | © 2012 Generation mPower LLC. All Rights Reserved. Reprinted with permission. The development of clean, affordable nuclear power options is a key element of the Department of Energy's Office of Nuclear Energy (DOE-NE) Nuclear Energy Research and Development Roadmap. As a part of this strategy, a high priority of the Department has been to help accelerate the timelines for the commercialization and deployment of small modular reactor (SMR) technologies through the SMR Licensing Technical Support program. Begun

256

Industry Spent Fuel Storage Handbook  

Science Conference Proceedings (OSTI)

The Industry Spent Fuel Storage Handbook (8220the Handbook8221) addresses the relevant aspects of at-reactor spent (or used) nuclear fuel (SNF) storage in the United States. With the prospect of SNF being stored at reactor sites for the foreseeable future, it is expected that all U.S. nuclear power plants will have to implement at-reactor dry storage by 2025 or shortly thereafter. The Handbook provides a broad overview of recent developments for storing SNF at U.S. reactor sites, focusing primarily on at...

2010-07-29T23:59:59.000Z

257

Nuclear reactor characteristics and operational history  

Gasoline and Diesel Fuel Update (EIA)

1. Capacity and Generation, Table 3. Characteristics and Operational History 1. Capacity and Generation, Table 3. Characteristics and Operational History Table 2. U.S. Nuclear Reactor Ownership Data PDF XLS Plant/Reactor Name Generator ID Utility Name - Operator Owner Name % Owned Arkansas Nuclear One 1 Entergy Arkansas Inc Entergy Arkansas Inc 100 Arkansas Nuclear One 2 Entergy Arkansas Inc Entergy Arkansas Inc 100 Beaver Valley 1 FirstEnergy Nuclear Operating Company FirstEnergy Nuclear Generation Corp 100 Beaver Valley 2 FirstEnergy Nuclear Operating Company FirstEnergy Nuclear Generation Corp 100 Braidwood Generation Station 1 Exelon Nuclear Exelon Nuclear 100 Braidwood Generation Station 2 Exelon Nuclear Exelon Nuclear 100 Browns Ferry 1 Tennessee Valley Authority Tennessee Valley Authority 100

258

NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY  

DOE Patents (OSTI)

A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

Stengel, F.G.

1963-12-24T23:59:59.000Z

259

Nuclear reactor composite fuel assembly  

DOE Patents (OSTI)

A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

Burgess, Donn M. (Richland, WA); Marr, Duane R. (West Richland, WA); Cappiello, Michael W. (Richland, WA); Omberg, Ronald P. (Richland, WA)

1980-01-01T23:59:59.000Z

260

Nuclear reactor control apparatus. [FBR  

DOE Patents (OSTI)

Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod.

Sridhar, B.N.

1981-04-16T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Digital computer operation of a nuclear reactor  

DOE Patents (OSTI)

A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

Colley, R.W.

1982-06-29T23:59:59.000Z

262

Liquid metal cooled nuclear reactor plant system  

DOE Patents (OSTI)

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

1993-01-01T23:59:59.000Z

263

Digital computer operation of a nuclear reactor  

DOE Patents (OSTI)

A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

Colley, Robert W. (Richland, WA)

1984-01-01T23:59:59.000Z

264

Disposition of ORNL's Spent Nuclear Fuel  

SciTech Connect

This paper describes the process of retrieving, repackaging, and preparing Oak Ridge spent nuclear fuel (SNF) for off-site disposition. The objective of the Oak Ridge SNF Project is to safely, reliably, and efficiently manage SNF that is stored on the Oak Ridge Reservation until it can be shipped off-site. The project required development of several unique processes and the design and fabrication of special equipment to enable the successful retrieval, transfer, and repackaging of Oak Ridge SNF. SNF was retrieved and transferred to a hot cell for repackaging. After retrieval of SNF packages, the storage positions were decontaminated and stainless steel liners were installed to resolve the vulnerability of water infiltration. Each repackaged SNF canister has been transferred from the hot cell back to dry storage until off-site shipments can be made. Three shipments of aluminum-clad SNF were made to the Savannah River Site (SRS), and five shipments of non-aluminum-clad SNF are planned to the Idaho National Engineering and Environmental Laboratory (INEEL). Through the integrated cooperation of several organizations including the U.S. Department of Energy (DOE), Bechtel Jacobs Company LLC (BJC), Oak Ridge National Laboratory (ORNL), and various subcontractors, preparations for the disposition of SNF in Oak Ridge have been performed in a safe and successful manner.

Turner, D. W.; DeMonia, B. C.; Horton, L. L.

2002-02-26T23:59:59.000Z

265

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

SciTech Connect

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

OLGUIN, L.J.

2000-09-25T23:59:59.000Z

266

Trojan Nuclear Plant Decommissioning: Final Survey for the Independent Spent Fuel Storage Installation Site  

Science Conference Proceedings (OSTI)

This report describes the final radiological survey for the area where Portland General Electric (PGE) will construct the Independent Spent Fuel Storage Installation (ISFSI) at Trojan nuclear power plant. The survey fulfills the requirements for release of this area from Trojan's 10 CFR 50 license before radiation levels increase with spent fuel storage in the ISFSI.

1998-05-13T23:59:59.000Z

267

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

SciTech Connect

This project developed methodology and field deployable tools (test kits) to analyze the chemical and microbiological condition of the fuel storage medium and determine the oxide thickness on the spent fuel basin materials. The overall objective of this project was to determine the amount of time fuel has spent in a storage basin to determine if the operation of the reactor and storage basin is consistent with safeguard declarations or expectations. This project developed and validated forensic tools that can be used to predict the age and condition of spent nuclear fuels stored in liquid basins based on key physical, chemical and microbiological basin characteristics. Key parameters were identified based on a literature review, the parameters were used to design test cells for corrosion analyses, tools were purchased to analyze the key parameters, and these were used to characterize an active spent fuel basin, the Savannah River Site (SRS) L-Area basin. The key parameters identified in the literature review included chloride concentration, conductivity, and total organic carbon level. Focus was also placed on aluminum based cladding because of their application to weapons production. The literature review was helpful in identifying important parameters, but relationships between these parameters and corrosion rates were not available. Bench scale test systems were designed, operated, harvested, and analyzed to determine corrosion relationships between water parameters and water conditions, chemistry and microbiological conditions. The data from the bench scale system indicated that corrosion rates were dependent on total organic carbon levels and chloride concentrations. The highest corrosion rates were observed in test cells amended with sediment, a large microbial inoculum and an organic carbon source. A complete characterization test kit was field tested to characterize the SRS L-Area spent fuel basin. The sampling kit consisted of a TOC analyzer, a YSI multiprobe, and a thickness probe. The tools were field tested to determine their ease of use, reliability, and determine the quality of data that each tool could provide. Characterization was done over a two day period in June 2011, and confirmed that the L Area basin is a well operated facility with low corrosion potential.

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

2012-06-04T23:59:59.000Z

268

Safer nuclear reactors could result from Los Alamos research  

NLE Websites -- All DOE Office Websites (Extended Search)

Calendar Video Newsroom News Releases News Releases - 2010 March Safer nuclear reactors could result from research Safer nuclear reactors could result from Los...

269

The role of actinide burning and the Integral Fast Reactor in the future of nuclear power  

Science Conference Proceedings (OSTI)

A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

1990-12-01T23:59:59.000Z

270

Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask  

Science Conference Proceedings (OSTI)

Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

Pope, R.B.; Diggs, J.M. (eds.)

1982-04-01T23:59:59.000Z

271

End-of-Life Rod Internal Pressures in Spent Pressurized Water Reactor Fuel  

Science Conference Proceedings (OSTI)

The end-of-reactor-life (EOL) rod internal pressure (RIP) is the primary protagonist for several evolutionary changes during long-term dry storage, which affect cladding resistance to failure when spent fuel assemblies are subjected to normal and accident conditions of transport. At the maximum temperature attained, either during vacuum drying or dry storage, EOL RIP determines the maximum stress state in the fuel rod cladding, which in turn sets the initial conditions for potential time-dependent ...

2013-12-17T23:59:59.000Z

272

Storage of LWR (light-water-reactor) spent fuel in air  

Science Conference Proceedings (OSTI)

An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to determine the oxidation response of light-water-reactor (LWR) spent fuels under conditions appropriate to fuel storage in air. The program is designed to investigate several independent variables that might affect the oxidation behavior of spent fuel. Included are temperature (135 to 230{degree}C), fuel burnup (to about 34 MWd/kgM), reactor type (pressurized and boiling water reactors), moisture level in the air, and the presence of a high gamma field. In continuing tests with declad spent fuel and nonirradiated UO{sub 2} specimens, oxidation rates were monitored by weight-gain measurements and the microstructures of subsamples taken during the weighing intervals were characterized by several analytical methods. The oxidation behavior indicated by weight gain and time to form powder will be reported in Volume III of this series. The characterization results obtained from x-ray diffractometry, transmission electron microscopy, scanning electron microscopy, and Auger electron spectrometry of oxidized fuel samples are presented in this report. 28 refs., 21 figs., 3 tabs.

Thomas, L.E.; Charlot, L.A.; Coleman, J.E. (Pacific Northwest Lab., Richland, WA (USA)); Knoll, R.W. (Johnson Controls, Inc., Madison, WI (USA))

1989-12-01T23:59:59.000Z

273

EIS-0250: Geologic Repository for the Disposal of Spent Nuclear Fuel and  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

EIS-0250: Geologic Repository for the Disposal of Spent Nuclear EIS-0250: Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada EIS-0250: Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada Summary This EIS analyzes DOE's proposed action to construct, operate, monitor, and eventually close a geologic repository at Yucca Mountain for the disposal of spent nuclear fuel and high-level radioactive waste. The EIS evaluates not only impacts from constructing, operating, monitoring, and closing a repository, but also from transporting the materials from 72 commercial and 4 DOE sites to the Yucca Mountain repository site in Nye County, Nevada. Public Comment Opportunities

274

Depleted Uraniuim Dioxide as a Spent-Nuclear-Fuel-Waste Package...  

NLE Websites -- All DOE Office Websites (Extended Search)

15 DEPLETED URANIUM DIOXIDE AS A SPENT-NUCLEAR-FUEL WASTE-PACKAGE PARTICULATE FILL: FILL BEHAVIOR Charles W. Forsberg Oak Ridge National Laboratory * P.O. Box 2008 Oak Ridge,...

275

EA-1117: Management of Spent Nuclear Fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee  

Energy.gov (U.S. Department of Energy (DOE))

This EA evaluates the environmental impacts of the proposal for the management of spent nuclear fuel on the U.S. Department of Energy's Oak Ridge Reservation to implement the preferred alternative...

276

How much spent (used) fuel is stored at U.S. nuclear power plants ...  

U.S. Energy Information Administration (EIA)

How much spent (used) fuel is stored at U.S. nuclear power plants? In 2002, the most recent year for which EIA has data, there were 161,662 fuel assemblies, or 46,268 ...

277

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

Science Conference Proceedings (OSTI)

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

MITCHELL, R.M.

2000-09-28T23:59:59.000Z

278

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

Science Conference Proceedings (OSTI)

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.

Koji Shirai

2006-04-01T23:59:59.000Z

279

Reactor Technology | Nuclear Science | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Research Areas Fuel Cycle Science & Technology Fusion Nuclear Science Isotope Development and Production Nuclear Security Science & Technology Nuclear Systems Modeling, Simulation...

280

Gas-cooled nuclear reactor  

DOE Patents (OSTI)

A gas-cooled nuclear reactor includes a central core located in the lower portion of a prestressed concrete reactor vessel. Primary coolant gas flows upward through the core and into four overlying heat-exchangers wherein stream is generated. During normal operation, the return flow of coolant is between the core and the vessel sidewall to a pair of motor-driven circulators located at about the bottom of the concrete pressure vessel. The circulators repressurize the gas coolant and return it back to the core through passageways in the underlying core structure. If during emergency conditions the primary circulators are no longer functioning, the decay heat is effectively removed from the core by means of natural convection circulation. The hot gas rising through the core exits the top of the shroud of the heat-exchangers and flows radially outward to the sidewall of the concrete pressure vessel. A metal liner covers the entire inside concrete surfaces of the concrete pressure vessel, and cooling tubes are welded to the exterior or concrete side of the metal liner. The gas coolant is in direct contact with the interior surface of the metal liner and transfers its heat through the metal liner to the liquid coolant flowing through the cooling tubes. The cooler gas is more dense and creates a downward convection flow in the region between the core and the sidewall until it reaches the bottom of the concrete pressure vessel when it flows radially inward and up into the core for another pass. Water is forced to flow through the cooling tubes to absorb heat from the core at a sufficient rate to remove enough of the decay heat created in the core to prevent overheating of the core or the vessel.

Peinado, Charles O. (La Jolla, CA); Koutz, Stanley L. (San Diego, CA)

1985-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

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

SciTech Connect

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

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

1995-01-31T23:59:59.000Z

282

Fast-acting nuclear reactor control device  

DOE Patents (OSTI)

A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

Kotlyar, Oleg M. (Idaho Falls, ID); West, Phillip B. (Idaho Falls, ID)

1993-01-01T23:59:59.000Z

283

Shutdown system for a nuclear reactor  

DOE Patents (OSTI)

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

1984-06-05T23:59:59.000Z

284

Shutdown system for a nuclear reactor  

DOE Patents (OSTI)

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

Groh, Edward F. (Naperville, IL); Olson, Arne P. (Western Springs, IL); Wade, David C. (Naperville, IL); Robinson, Bryan W. (Oak Lawn, IL)

1984-01-01T23:59:59.000Z

285

Shutdown system for a nuclear reactor  

DOE Patents (OSTI)

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

1982-01-20T23:59:59.000Z

286

Application of ALARA principles to shipment of spent nuclear fuel  

SciTech Connect

The public exposure from spent fuel shipment is very low. In view of this low exposure and the perfect safety record for spent fuel shipment, existing systems can be considered satisfactory. On the other hand, occupational exposure reduction merits consideration and technology improvement to decrease dose should concentrate on this exposure. Practices that affect the age of spent fuel in shipment and the number of times the fuel must be shipped prior to disposal have the largest impact. A policy to encourage a 5-year spent fuel cooling period prior to shipment coupled with appropriate cask redesign to accommodate larger loads would be consistent with ALARA and economic principles. And finally, bypassing high population density areas will not in general reduce shipment dose.

Greenborg, J.; Brackenbush, L.W.; Murphy, D.W. Burnett, R.A.; Lewis, J.R.

1980-05-01T23:59:59.000Z

287

Method for storing spent nuclear fuel in repositories  

DOE Patents (OSTI)

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.

Schweitzer, Donald G. (Bayport, NY); Sastre, Cesar (Shoreham, NY); Winsche, Warren (Bellport, NY)

1981-01-01T23:59:59.000Z

288

EIS-0218: Proposed Nuclear Weapons Nonproliferation Policy Concerning...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

18: Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel EIS-0218: Proposed Nuclear Weapons Nonproliferation Policy Concerning...

289

Cost probability analysis of reprocessing spent nuclear fuel in the US G.D. Recktenwald, M.R. Deinert  

E-Print Network (OSTI)

to the sustainability of nuclear power while others argue against them on economic, environmental and security groundsCost probability analysis of reprocessing spent nuclear fuel in the US G.D. Recktenwald, M P48 Keywords: Reprocessing Nuclear power Spent fuel The methods by which nuclear power's radioactive

Deinert, Mark

290

Nuclear propulsion apparatus with alternate reactor segments  

DOE Patents (OSTI)

1. Nuclear propulsion apparatus comprising: A. means for compressing incoming air; B. nuclear fission reactor means for heating said air; C. means for expanding a portion of the heated air to drive said compressing means; D. said nuclear fission reactor means being divided into a plurality of radially extending segments; E. means for directing a portion of the compressed air for heating through alternate segments of said reactor means and another portion of the compressed air for heating through the remaining segments of said reactor means; and F. means for further expanding the heated air from said drive means and the remaining heated air from said reactor means through nozzle means to effect reactive thrust on said apparatus.

Szekely, Thomas (Santa Monica, CA)

1979-04-03T23:59:59.000Z

291

BOOK: Safety Related Issues of Spent Nuclear Fuel Storage  

Science Conference Proceedings (OSTI)

Sep 26, 2007... Trends in Nuclear Power, The Nuclear Fuel Cycle, Nuclear Science ... Fifteen papers cover aluminum-clad fuel discharged from research ...

292

Large Scale Weather Control Using Nuclear Reactors  

E-Print Network (OSTI)

It is pointed out that controlled release of thermal energy from fission type nuclear reactors can be used to alter weather patterns over significantly large geographical regions. (1) Nuclear heat creates a low pressure region, which can be used to draw moist air from oceans, onto deserts. (2) Creation of low pressure zones over oceans using Nuclear heat can lead to Controlled Cyclone Creation (CCC).(3) Nuclear heat can also be used to melt glaciers and control water flow in rivers.

Moninder Singh Modgil

2002-10-02T23:59:59.000Z

293

Nuclear reactor shield including magnesium oxide  

DOE Patents (OSTI)

An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

Rouse, Carl A. (Del Mar, CA); Simnad, Massoud T. (La Jolla, CA)

1981-01-01T23:59:59.000Z

294

Fuel handling apparatus for a nuclear reactor  

DOE Patents (OSTI)

Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

Hawke, Basil C. (Solana Beach, CA)

1987-01-01T23:59:59.000Z

295

Today and Future Neutrino Experiments at Krasnoyarsk Nuclear Reactor  

E-Print Network (OSTI)

The results of undergoing experiments and new experiment propositions at Krasnoyarsk underground nuclear reactor are presented

Yu. V. Kozlov; S. V. Khalturtsev; I. N. Machulin; A. V. Martemyanov; V. P. Martemyanov; A. A. Sabelnikov; V. G. Tarasenkov; E. V. Turbin; V. N. Vyrodov; L. A. Popeko; A. V. Cherny; G. A. Shishkina

1999-12-21T23:59:59.000Z

296

MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

Balent, R.

1963-03-12T23:59:59.000Z

297

Dynamic detection of nuclear reactor core incident  

Science Conference Proceedings (OSTI)

Surveillance, safety and security of evolving systems are a challenge to prevent accident. The dynamic detection of a hypothetical and theoretical blockage incident in the Phenix nuclear reactor is investigated. Such an incident is characterized by abnormal ... Keywords: Contrast, Dynamic detection of perturbations, Evolving system, Fast-neutron reactor, Neighbourhood, Noise

Laurent Hartert; Danielle Nuzillard; Jean-Philippe Jeannot

2013-02-01T23:59:59.000Z

298

Nuclear reactor vessel fuel thermal insulating barrier  

DOE Patents (OSTI)

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

2013-03-19T23:59:59.000Z

299

Early Argonne reactor lit the way for worldwide nuclear industry -  

NLE Websites -- All DOE Office Websites (Extended Search)

Early Argonne reactor lit the way for worldwide Early Argonne reactor lit the way for worldwide nuclear industry About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

300

Computer simulations help design new nuclear reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Computer simulations help design new nuclear reactors Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear Reactors Designed by Argonne Argonne's Nuclear Science and Technology Legacy Opportunities within NE Division Visit Argonne Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Highlights Bookmark and Share Reprinted from "Argonne Now" - Spring 2008 Physicist Won-Sik Yang and computer scientist Andrew Siegel hold a fuel rod assembly in front of a model of the Experimental Breeder Reactor-II

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Why Nuclear Energy? - Reactors designed/built by Argonne National  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Energy: Nuclear Energy: Why Nuclear Energy? About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

302

High-Fidelity Light Water Reactor Analysis with the Numerical Nuclear Reactor  

Science Conference Proceedings (OSTI)

Technical Paper / Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications

David P. Weber; Tanju Sofu; Won Sik Yang; Thomas J. Downar; Justin W. Thomas; Zhaopeng Zhong; Jin Young Cho; Kang Seog Kim; Tae Hyun Chun; Han Gyu Joo; Chang Hyo Kim

303

Nuclide Composition Benchmark Data Set for Verifying Burnup Codes on Spent Light Water Reactor Fuels  

SciTech Connect

To establish a nuclide composition benchmark data set for the verification of burnup codes, destructive analyses of light water reactor spent-fuel samples, which were cut out from several heights of spent-fuel rods, were carried out at the analytical laboratory at the Japan Atomic Energy Research Institute. The 16 samples from three kinds of pressurized water reactor (PWR) fuel rods and the 18 samples from two boiling water reactor (BWR) fuel rods were examined. Their initial {sup 235}U enrichments and burnups were from 2.6 to 4.1% and from 4 to 50 GWd/t, respectively. One PWR fuel rod and one BWR fuel rod contained gadolinia as a burnable poison. The measurements for more than 40 nuclides of uranium, transuranium, and fission product elements were performed by destructive analysis using mass spectrometry, and alpha-ray and gamma-ray spectrometry. Burnup for each sample was determined by the {sup 148}Nd method. The analytical methods and the results as well as the related irradiation condition data are compiled as a complete benchmark data set.

Nakahara, Yoshinori; Suyama, Kenya; Inagawa, Jun; Nagaishi, Ryuji; Kurosawa, Setsumi; Kohno, Nobuaki; Onuki, Mamoru; Mochizuki, Hiroki [Japan Atomic Energy Research Institute (Japan)

2002-02-15T23:59:59.000Z

304

Heat dissipating nuclear reactor with metal liner  

DOE Patents (OSTI)

Disclosed is a nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

Gluekler, Emil L. (San Jose, CA); Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

1987-01-01T23:59:59.000Z

305

Heat dissipating nuclear reactor with metal liner  

DOE Patents (OSTI)

A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

1985-11-21T23:59:59.000Z

306

Reactivity control assembly for nuclear reactor. [LMFBR  

DOE Patents (OSTI)

This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

Bollinger, L.R.

1982-03-17T23:59:59.000Z

307

Nuclear reactor construction with bottom supported reactor vessel  

DOE Patents (OSTI)

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

Sharbaugh, John E. (Bullskin Township, Fayette County, PA)

1987-01-01T23:59:59.000Z

308

Separator assembly for use in spent-nuclear-fuel shipping cask. [Patent application  

DOE Patents (OSTI)

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.

Bucholz, J.A.

1981-04-24T23:59:59.000Z

309

High-Energy Delayed Gamma Spectroscopy for Spent Nuclear Fuel Assay  

SciTech Connect

High-accuracy, direct, nondestructive measurement of fissile and fissionable isotopes in spent fuel, particularly the Pu isotopes, is a well-documented, but still unmet challenge in international safeguards. As nuclear fuel cycles propagate around the globe, the need for improved materials accountancy techniques for irradiated light-water reactor fuel will increase. This modeling study investigates the use of delayed gamma rays from fission-product nuclei to directly measure the relative concentrations of U-235, Pu-239, and Pu-241 in spent fuel assemblies. The method is based on the unique distribution of fission-product nuclei produced from fission in each of these fissile isotopes. Fission is stimulated in the assembly with a pulse-capable source of interrogating neutrons. The measured distributions of the short-lived fission products from the unknown sample are then fit with a linear combination of the known fission-product yield curves from pure U-235, Pu-239, and Pu-241 to determine the original proportions of these fissile isotopes. Modeling approaches for the intense gamma-ray background promulgated by the long-lived fission-product inventory and for the high-energy gamma-ray signatures emitted by short-lived fission products from induced fission are described. Benchmarking measurements are presented and compare favorably with the results of these models. Results for the simulated assay of simplified individual fuel elements ranging from fresh to 60 GWd/MTU burnup demonstrate the utility of the modeling methods for viability studies, although additional work is needed to more realistically assess the potential of High-Energy Delayed Gamma Spectroscopy (HEDGS).

Campbell, Luke W.; Smith, Leon E.; Misner, Alex C.

2011-02-01T23:59:59.000Z

310

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

Science Conference Proceedings (OSTI)

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

NONE

1996-07-01T23:59:59.000Z

311

Structural Integrity of Advanced Claddings During Spent Nuclear Fuel Transportation and Storage  

Science Conference Proceedings (OSTI)

Thermal creep is the dominant deformation mechanism of fuel cladding during transportation and dry storage of spent nuclear fuel. Thermal creep data and creep models of Westinghouse ZIRLO and LK3 cladding tubes were generated for use in spent-fuel storage and transportation applications. The final report consists of two volumes. This document (Volume 1) provides the project results obtained on non-irradiated and irradiated standard ZIRLO and non-irradiated optimized ZIRLO claddings.

2011-06-28T23:59:59.000Z

312

Impacts Associated with Transfer of Spent Nuclear Fuel from Spent Fuel Storage Pools to Dry Storage After Five Years of Cooling, Revision 1  

Science Conference Proceedings (OSTI)

In 2010, EPRI performed a study of the accelerated transfer of spent fuel from pools to dry storage in response to the threat of terrorist activities at nuclear power plants (report 1021049). Following the March 2011 Great East Japan Earthquake and the subsequent accident at the Fukushima Daiichi nuclear power plant, some organizations issued a renewed call for accelerated transfer of used fuel from spent fuel ...

2012-08-31T23:59:59.000Z

313

IMPROVEMENTS RELATING TO NUCLEAR REACTORS  

SciTech Connect

In order to reduce the pumping power for the coolant in a steam-cooled reactor, in which the steam being passed through successive sections of the reactor core and being superheated there, the sections are connected in series with one another, while a plurality of de-superheaters is provided such that steam flowing from one section to the next passes through a de-superheater. The condensed steam returning to the reactor from the means utilizing the steam heat content is divided into a number of separate streams. The first stream going to the first section in the reactor core is raised at least to saturated steam outside the reactor, while the remaining streams of condensed steam are conveyed to the de-superheaters to be mixed with steam passing therethrough between successive sections of the reactor, cooling in this manner said steam and being themselves converted into steam. Increasing amounts of condensate are added in successive de-superheaters until the steam returning to the reactor from the final desuperheater is equivalent to the full mass flow of steam circulating to the heat utilizing means. (NPO)

1960-08-01T23:59:59.000Z

314

On selection and operation of an international interim storage facility for spent nuclear fuel  

E-Print Network (OSTI)

Disposal of post-irradiation fuel from nuclear reactors has been an issue for the nuclear industry for many years. Most countries currently have no long-term disposal strategy in place. Therefore, the concept of an ...

Burns, Joe, 1966-

2004-01-01T23:59:59.000Z

315

ME 361E Nuclear Reactor Engineering ABET EC2000 syllabus  

E-Print Network (OSTI)

ME 361E ­ Nuclear Reactor Engineering Page 1 ABET EC2000 syllabus ME 361E ­ Nuclear Reactor; neutron diffusion and moderation; reactor equations; Fermi Age theory; multigroup and multiregional students should be able to: · Compare and contrast numerous nuclear reactor designs · Calculate the effects

Ben-Yakar, Adela

316

STEAM GENERATOR FOR NUCLEAR REACTOR  

DOE Patents (OSTI)

The steam generator described for use in reactor powergenerating systems employs a series of concentric tubes providing annular passage of steam and water and includes a unique arrangement for separating the steam from the water. (AEC)

Kinyon, B.W.; Whitman, G.D.

1963-07-16T23:59:59.000Z

317

TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES  

Science Conference Proceedings (OSTI)

A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading. A measurement of the average assembly burnup is required and that measurement must be within 10% of the utility burnup record for the assembly to be accepted. The measurement device must be accurate to within 10%. Each step is described in detail for use with any computer code system and is then demonstrated with the SCALE 4.2 computer code package using 27BURNUPLIB cross sections.

DOE

1997-04-01T23:59:59.000Z

318

MOLTEN FLUORIDE NUCLEAR REACTOR FUEL  

DOE Patents (OSTI)

Molten-salt reactor fuel compositions consisting of mixtures of fluoride salts are reported. In its broadest form, the composition contains an alkali fluoride such as sodium fluoride, zirconium tetrafluoride, and a uranium fluoride, the latter being the tetrafluoride or trifluoride or a mixture of the two. An outstanding property of these fuel compositions is a high coeffieient of thermal expansion which provides a negative temperature coefficient of reactivity in reactors in which they are used.

Barton, C.J.; Grimes, W.R.

1960-01-01T23:59:59.000Z

319

Optimally moderated nuclear fission reactor and fuel source therefor  

DOE Patents (OSTI)

An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

Ougouag, Abderrafi M. (Idaho Falls, ID); Terry, William K. (Shelley, ID); Gougar, Hans D. (Idaho Falls, ID)

2008-07-22T23:59:59.000Z

320

Quantifying the passive gamma signal from spent nuclear fuel in support of determining the plutonium content in spent nuclear fuel with nondestructive assay  

SciTech Connect

The objective of safeguarding nuclear material is to deter diversions of significant quantities of nuclear materials by timely monitoring and detection. There are a variety of motivations for quantifying plutonium in spent fuel (SF), by means of nondestructive assay (NDA), in order to meet this goal. These motivations include the following: strengthening the capabilities of the International Atomic Energy Agencies ability to safeguard nuclear facilities, shipper/receiver difference, input accountability at reprocessing facilities and burnup credit at repositories. Many NDA techniques exist for measuring signatures from SF; however, no single NDA technique can, in isolation, quantify elemental plutonium in SF. A study has been undertaken to determine the best integrated combination of 13 NDA techniques for characterizing Pu mass in spent fuel. This paper focuses on the development of a passive gamma measurement system in support the spent fuel assay system. Gamma ray detection for fresh nuclear fuel focuses on gamma ray emissions that directly coincide with the actinides of interest to the assay. For example, the 186-keV gamma ray is generally used for {sup 235}U assay and the 384-keV complex is generally used for assaying plutonium. In spent nuclear fuel, these signatures cannot be detected as the Compton continuum created from the fission products dominates the signal in this energy range. For SF, the measured gamma signatures from key fission products ({sup 134}Cs, {sup 137}Cs, {sup 154}Eu) are used to ascertain burnup, cooling time, and fissile content information. In this paper the Monte Carlo modeling set-up for a passive gamma spent fuel assay system will be described. The set-up of the system includes a germanium detector and an ion chamber and will be used to gain passive gamma information that will be integrated into a system for determining Pu in SF. The passive gamma signal will be determined from a library of {approx} 100 assemblies that have been created to examine the capability of all 13 NDA techniques. Presented in this paper is a description of the passive gamma monitoring instrument, explanation of the work completed thus far involving the source set up methodology and the design optimization process, details of key fission product ratios of interest, limitations and key strengths of the measurement technique, and considerations for integrating this technique with other NDA techniques in order to develop a complete spent fuel assay strategy.

Fensin, Michael L [Los Alamos National Laboratory; Tobin, Steven J [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory

2009-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Structural mechanics of fast spectrum nuclear reactor cores  

NLE Websites -- All DOE Office Websites (Extended Search)

mechanics of fast spectrum nuclear reactor cores A fast reactor core is composed of a closely packed hexagonal arrangement of fuel, control, blanket , and shielding assemblies....

322

Export possibilities for small nuclear reactors  

Science Conference Proceedings (OSTI)

The worldwide deployment of peaceful nuclear technology is predicated on conformance with the Nuclear Non-Proliferation Treaty of 1972. Under this international treaty, countries have traded away pursuit of nuclear weapons in exchange for access to commercial nuclear technology that could help them grow economically. Realistically, however, most nuclear technology has been beyond the capacity of the NPT developing countries to afford. Even if the capital cost of the plant is managed, the costs of the infrastructure and the operational complexity of most nuclear technology have taken it out of the hands of the nations who need it the most. Now, a new class of small sodium cooled reactors has been specifically designed to meet the electrical power, water, hydrogen and heat needs of small and remote users. These reactors feature small size, long refueling interval, no onsite fuel storage, and simplified operations. Sized in the 10 MW(e) to 50 MW(e) range these reactors are modularized for factory production and for rapid site assembly. The fuel would be <20% U-235 uranium fuel with a 30-year core life. This new reactor type more appropriately fills the needs of countries for lower power distributed systems that can fill the gap between large developed infrastructure and primitive distributed energy systems. Looking at UN Resolution 1540 and the impact of other agreements, there is a need to address the issues of nuclear security, fuel, waste, and economic/legal/political-stakeholder concerns. This paper describes the design features of this new reactor type that specifically address these issues in a manner that increases the availability of commercial nuclear technology to the developing nations of the world. (authors)

Campagna, M.S.; Hess, C.; Moor, P. [Burns and Roe Enterprises, Inc., Oradell, NJ (United States); Sawruk, W. [ABSG Consulting, Inc., Shillington, PA (United States)

2007-07-01T23:59:59.000Z

323

White paper report on using nuclear reactors to search for a value of theta13  

E-Print Network (OSTI)

PAPER REPORT on Using Nuclear Reactors to Search for a valuetimely new experiment at a nuclear reactor sensitive to theand judicious choice of a nuclear reactor. The dominant

2004-01-01T23:59:59.000Z

324

Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors  

E-Print Network (OSTI)

L. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,R. J. Neuhold, Introductury Nuclear Reactor Dynamics. ANSL. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,

Galvez, Cristhian

2011-01-01T23:59:59.000Z

325

Integrated Data Base report--1993: U.S. spent nuclear fuel and radioactive waste inventories, projections, and characteristics. Revision 10  

Science Conference Proceedings (OSTI)

The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and DOE spent nuclear fuel; also, commercial and US government-owned radioactive wastes through December 31, 1993. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 256 refs., 38 figs., 141 tabs.

Not Available

1994-12-01T23:59:59.000Z

326

Cooling system for a nuclear reactor  

DOE Patents (OSTI)

A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

Amtmann, Hans H. (Rancho Santa Fe, CA)

1982-01-01T23:59:59.000Z

327

Fast-acting nuclear reactor control device  

DOE Patents (OSTI)

This invention consists of a fast-acting nuclear reactor control device for moving and positioning a safety control rod to desired elevations within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump motor, an electric gear motor, and a solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch, allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

Kotlyar, O.M.; West, P.B.

1992-12-31T23:59:59.000Z

328

Nuclear reactor fissile isotopes antineutrino spectra  

E-Print Network (OSTI)

Positron spectrum from inverse beta decay reaction on proton was measured in 1988-1990 as a result of neutrino exploration experiment. The measured spectrum has the largest statistics and lowest energy threshold between other neutrino experiments made that time at nuclear reactors. On base of the positron spectrum the standard antineutrino spectrum for typical reactor fuel composition was restored. In presented analysis the partial spectra forming this standard spectrum were extracted using specific method. They could be used for neutrino experiments data analysis made at any fuel composition of reactor core.

Sinev, V

2012-01-01T23:59:59.000Z

329

Nuclear reactor fissile isotopes antineutrino spectra  

E-Print Network (OSTI)

Positron spectrum from inverse beta decay reaction on proton was measured in 1988-1990 as a result of neutrino exploration experiment. The measured spectrum has the largest statistics and lowest energy threshold between other neutrino experiments made that time at nuclear reactors. On base of the positron spectrum the standard antineutrino spectrum for typical reactor fuel composition was restored. In presented analysis the partial spectra forming this standard spectrum were extracted using specific method. They could be used for neutrino experiments data analysis made at any fuel composition of reactor core.

V. Sinev

2012-07-30T23:59:59.000Z

330

Theta 13 Determination with Nuclear Reactors  

E-Print Network (OSTI)

Recently there has been a lot of interest around the world in the use of nuclear reactors to measure theta 13, the last undetermined angle in the 3-neutrino mixing scenario. In this paper the motivations for theta 13 measurement using short baseline nuclear reactor experiments are discussed. The features of such an experiment are described in the context of Double Chooz, which is a new project planned to start data-taking in 2008, and to reach a sensitivity of sinsq(2 theta 13) < 0.03.

F. Dalnoki-Veress

2004-06-24T23:59:59.000Z

331

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

Science Conference Proceedings (OSTI)

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

Not Available

1994-06-01T23:59:59.000Z

332

Spent nuclear fuel storage -- Performance tests and demonstrations  

SciTech Connect

This report summarizes the results of heat transfer and shielding performance tests and demonstrations conducted from 1983 through 1992 by or in cooperation with the US Department of Energy (DOE), Office of Commercial Radioactive Waste Management (OCRWM). The performance tests consisted of 6 to 14 runs involving one or two loadings, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. A description of the test plan, spent fuel load patterns, results from temperature and dose rate measurements, and fuel integrity evaluations are contained within the report.

McKinnon, M.A.; DeLoach, V.A.

1993-04-01T23:59:59.000Z

333

Spent Nuclear Fuel (SNF) Project Safety Basis Implementation Strategy  

Science Conference Proceedings (OSTI)

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

TRAWINSKI, B.J.

2000-02-08T23:59:59.000Z

334

Characteristics of Spent Fuel from Plutonium Disposition Reactors. Vol. 3: A Westinghouse Pressurized-Water Reactor Design  

Science Conference Proceedings (OSTI)

This report discusses the results of a simulation study involving the burnup of mixed-oxide (MOX) fuel in a Westinghouse pressurized-water reactor (PWR). The MOX was composed of uranium and plutonium oxides, where the plutonium was of weapons-grade composition. The study was part of the Fissile Materials Disposition Program and considered the possibility of fueling commercial reactors with weapons plutonium. The isotopic composition, the activities, and the decay heat, together with the gamma and neutron dose rates are discussed for the spent fuel. For the steady-state situation involving this PWR burning MOX fuel, two burn histories are reported. In one case, an assembly is burned in the reactor for two cycles, and in the second case and assembly is burned for three cycles. Furthermore, assemblies containing wet annular burnable absorbers (WABAs) and assemblies that do not contain WABAs are considered in all cases. The two-cycle cases have a burnup of 35 GWd/t, and the three-cycle cases have a burnup of 52.5 GWd/t.

Murphy, B.D.

1997-07-01T23:59:59.000Z

335

Office of Nuclear Reactor Regulation  

E-Print Network (OSTI)

The U.S. Nuclear Regulatory Commission (NRC) is considering renewal of the operating licenses for the Edwin I. Hatch Nuclear Plant, Units 1 and 2 (HNP) for a period of an additional 20 years. The purpose of this assessment is to provide information to the U.S. National Marine Fisheries Service concerning the impacts of continued operation of the HNP on the shortnose sturgeon, Acipenser brevirostrum. The

unknown authors

2000-01-01T23:59:59.000Z

336

Technological Transfer from Research Nuclear Reactors to New Generation Nuclear Power Reactors  

SciTech Connect

The goal of this paper is the analysis of the technological transfer role in the nuclear field, with particular emphasis on nuclear reactors domain. The presentation is sustained by historical arguments. In this frame, it is very important to start with the achievements of the first nuclear systems, for instant those with natural uranium as fuel and heavy water as moderator, following in time through the history until the New Generation Nuclear Power Reactors.Starting with 1940, the accelerated development of the industry has implied the increase of the global demand for energy. In this respect, the nuclear energy could play an important role, being essentially an unlimited source of energy. However, the nuclear option faces the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide, a significant amount of experience has been accumulated during development, licensing, construction, and operation of nuclear power reactors. The experience gained is a strong basis for further improvements. Actually, the nuclear programs of many countries are addressing the development of advanced reactors, which are intended to have better economics, higher reliability, improved safety, and proliferation-resistant characteristics in order to overcome the current concerns about nuclear power. Advanced reactors, now under development, may help to meet the demand for energy power of both developed and developing countries as well as for district heating, desalination and for process heat.The paper gives historical examples that illustrate the steps pursued from first research nuclear reactors to present advanced power reactors. Emphasis was laid upon the fact that the progress is due to the great discoveries of the nuclear scientists using the technological transfer.

Radulescu, Laura ['Horia Hulubei' National Institute of Nuclear Physics and Engineering, PO BOX MG-6, Bucharest 077125 (Romania); Pavelescu, Margarit [Academy of Romanian Scientists, Bucharest (Romania)

2010-01-21T23:59:59.000Z

337

22.312 Engineering of Nuclear Reactors, Fall 2004  

E-Print Network (OSTI)

Engineering principles of nuclear reactors, emphasizing power reactors. Power plant thermodynamics, reactor heat generation and removal (single-phase as well as two-phase coolant flow and heat transfer), and structural ...

Buongiorno, Jacopo, 1971-

338

22.312 Engineering of Nuclear Reactors, Fall 2002  

E-Print Network (OSTI)

Engineering principles of nuclear reactors, emphasizing power reactors. Power plant thermodynamics, reactor heat generation and removal (single-phase as well as two-phase coolant flow and heat transfer), and structural ...

Todreas, Neil E.

339

Basis for Identification of Disposal Options for R and D for Spent Nuclear  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Basis for Identification of Disposal Options for R and D for Spent Basis for Identification of Disposal Options for R and D for Spent Nuclear Fuel and High-Level Waste Basis for Identification of Disposal Options for R and D for Spent Nuclear Fuel and High-Level Waste The Used Fuel Disposition campaign (UFD) is selecting a set of geologic media for further study that spans a suite of behavior characteristics that impose a broad range of potential conditions on the design of the repository, the engineered barrier, and the waste. Salt, clay/shale, and granitic rocks represent a reasonable cross-section of behavior. Granitic rocks are also the primary basement rock to consider for deep borehole disposal. UFD is developing generic system analysis capability and general experimental data related to mined geologic disposal in the three

340

Spent graphite fuel element processing  

SciTech Connect

The Department of Energy currently sponsors two programs to demonstrate the processing of spent graphite fuel elements. General Atomic in San Diego operates a cold pilot plant to demonstrate the processing of both US and German high-temperature reactor fuel. Exxon Nuclear Idaho Company is demonstrating the processing of spent graphite fuel elements from Rover reactors operated for the Nuclear Rocket Propulsion Program. This work is done at Idaho National Engineering Laboratory, where a hot facility is being constructed to complete processing of the Rover fuel. This paper focuses on the graphite combustion process common to both programs.

Holder, N.D.; Olsen, C.W.

1981-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Investigation of bond graphs for nuclear reactor simulations  

E-Print Network (OSTI)

This work proposes a simple and effective approach to modeling multiphysics nuclear reactor problems using bond graphs. The conventional method of modeling the coupled multiphysics transients in nuclear reactors is operator ...

Sosnovsky, Eugeny

2010-01-01T23:59:59.000Z

342

More About NNSA's Naval Reactors Office | National Nuclear Security...  

National Nuclear Security Administration (NNSA)

Naval Reactors Office The Naval Nuclear Propulsion Program provides militarily effective nuclear propulsion plants and ensures their safe, reliable and long-lived operation. This...

343

Domestic U.S. Reactor Conversions: Fact Sheet | National Nuclear...  

NLE Websites -- All DOE Office Websites (Extended Search)

U.S. Reactor Conversions: Fact Sheet | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency...

344

Damper mechanism for nuclear reactor control elements  

DOE Patents (OSTI)

A damper mechanism which provides a nuclear reactor control element decelerating function at the end of the scram stroke. The total damping function is produced by the combination of two assemblies, which operate in sequence. First, a tapered dashram assembly decelerates the control element to a lower velocity, after which a spring hydraulic damper assembly takes over to complete the final damping.

Taft, William Elwood (Los Gatos, CA)

1976-01-01T23:59:59.000Z

345

Current Abstracts Nuclear Reactors and Technology  

SciTech Connect

This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

Bales, J.D.; Hicks, S.C. [eds.

1993-01-01T23:59:59.000Z

346

NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT  

DOE Patents (OSTI)

A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

Currier, E.L. Jr.; Nicklas, J.H.

1962-08-14T23:59:59.000Z

347

Nuclear reactor shutdown control rod assembly  

DOE Patents (OSTI)

A temperature responsive, self-actuated nuclear reactor shutdown control rod assembly 10. The upper end 18 of a lower drive line 17 fits within the lower end of an upper drive line 12. The lower end (not shown) of the lower drive line 17 is connected to a neutron absorber. During normal temperature conditions the lower drive line 17 is supported by detent means 22,26. When an overtemperature condition occurs thermal actuation means 34 urges ring 26 upwardly sufficiently to allow balls 22 to move radially outwardly thereby allowing lower drive line 17 to move downwardly toward the core of the nuclear reactor resulting in automatic reduction of the reactor powder.

Bilibin, Konstantin (North Hollywood, CA)

1988-01-01T23:59:59.000Z

348

Spent Nuclear Fuel project photon heat deposition calculation for hygrogen generation within MCO  

DOE Green Energy (OSTI)

Three types of water conditions are analyzed for nuclear heat deposition in a MCO: fully flooded, thick film, and thin film. These heat deposition rates within water can be used to determine gas generation during the different phases of Spent Fuel removal and processing for storage.

Lan, J.S.

1996-08-01T23:59:59.000Z

349

A safety and dynamics analysis of the subcritical advanced burner reactor: SABR.  

E-Print Network (OSTI)

??As the United States expands its quantity of nuclear reactors in the near future, the amount of spent nuclear fuel (SNF) will also increase. Closing… (more)

Sumner, Tyler Scott

2008-01-01T23:59:59.000Z

350

Characterization of Nuclear Reactor Materials and Components with ...  

Science Conference Proceedings (OSTI)

About this Symposium. Meeting, 2013 TMS Annual Meeting & Exhibition. Symposium, Characterization of Nuclear Reactor Materials and Components with  ...

351

Characterization of Nuclear Reactor Materials and Components with ...  

Science Conference Proceedings (OSTI)

About this Symposium. Meeting, 2011 TMS Annual Meeting & Exhibition. Symposium, Characterization of Nuclear Reactor Materials and Components with  ...

352

Heat pipe nuclear reactor for space power  

SciTech Connect

A heat-pipe cooled nuclear reactor has been designed to provide 3.2 MW(t) to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat pipe temperature of 1675/sup 0/K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum, lithium vapor, heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO/sub 2/ pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber and a BeO reflector containing boron loaded control drums.

Koenig, D.R.

1976-01-01T23:59:59.000Z

353

Passive heat transfer means for nuclear reactors  

DOE Patents (OSTI)

An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. Means such as shrouding normally isolated the secondary condensing section from effective heat transfer with the heat sink, but a sensor responds to overheat conditions of the reactor to open the shrouding, which thereby increases the cooling capacity of the heat pipe. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

Burelbach, James P. (Glen Ellyn, IL)

1984-01-01T23:59:59.000Z

354

Depleted uranium oxides as spent-nuclear-fuel waste-package invert and backfill materials  

SciTech Connect

A new technology has been proposed in which depleted uranium, in the form of oxides or silicates, is placed around the outside of the spent nuclear fuel waste packages in the geological repository. This concept may (1) reduce the potential for repository nuclear criticality events and (2) reduce long-term release of radionuclides from the repository. As a new concept, there are significant uncertainties.

Forsberg, C.W.; Haire, M.J.

1997-07-07T23:59:59.000Z

355

MA50177: Scientific Computing Nuclear Reactor Simulation Generalised Eigenvalue Problems  

E-Print Network (OSTI)

MA50177: Scientific Computing Case Study Nuclear Reactor Simulation ­ Generalised Eigenvalue of a malfunction or of an accident experimentally, the numerical simulation of nuclear reactors is of utmost balance in a nuclear reactor are the two-group neutron diffusion equations -div (K1 u1) + (a,1 + s) u1 = 1

Scheichl, Robert

356

1 INTRODUCTION Modern nuclear reactor concepts make use of pas-  

E-Print Network (OSTI)

1 INTRODUCTION Modern nuclear reactor concepts make use of pas- sive safety features (Fong et al systems in advanced nuclear reactors; in (Cardoso et al. 2008), Artificial Neural Networks (ANNs: Special Issue "Natural Circulation in Nuclear Reactor Systems", Hindawi Publishing Corpo- ration, Paper

357

Polynomial regression with derivative information in nuclear reactor uncertainty quantification*  

E-Print Network (OSTI)

1 Polynomial regression with derivative information in nuclear reactor uncertainty quantification in the outputs. The usual difficulties in modeling the work of the nuclear reactor models include the large size, applying the existing AD tools to nuclear reactor models still takes considerable development effort

Anitescu, Mihai

358

LIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS  

E-Print Network (OSTI)

of a nuclear reactor with feedback," in: Applied Problems in the Theory of Oscillations [in RussianLIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS M. V. Bazhenov and E. F. Sabaev UDC of Nuclear Reactors [in Russian], l~nergoatomizdat, Moscow (1990). F. R. Gantmakher and V. A. Yakubovich

Bazhenov, Maxim

359

FUEL ELEMENT FOR NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear fuel element comprising a large number og wafers of fissionable material and a protective jacket having compartments holding these wafers is described. The compartments of the jacket aid the removal of heat from the wafers, keep the wafers or fragments thereof from migrating in the jacket, and permit the escape of gaseous fission products.

Carney, K.G. Jr.

1959-07-14T23:59:59.000Z

360

DECOMMISSIONING OF NUCLEAR POWER REACTORS  

E-Print Network (OSTI)

Decommissioning means permanently removing a nuclear facility from service and reducing radioactive material on the licensed site to levels that would permit termination of the NRC license. On June 27, 1988, the NRC issued general requirements on decommissioning that contained technical and financial criteria and dealt with planning needs, timing, funding mechanisms, and environmental review

unknown authors

2000-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Nuclear & Uranium - Analysis & Projections - U.S. Energy ...  

U.S. Energy Information Administration (EIA)

Uranium fuel, nuclear reactors, generation, spent fuel. Total Energy. ... Nuclear power plants generate approximately 20 percent of U.S. electricity, ...

362

Spent fuel characteristics & disposal considerations  

SciTech Connect

The fuel used in commercial nuclear power reactors is uranium, generally in the form of an oxide. The gas-cooled reactors developed in England use metallic uranium enclosed in a thin layer of Magnox. Since this fuel must be processed into a more stable form before disposal, we will not consider the characteristics of the Magnox spent fuel. The vast majority of the remaining power reactors in the world use uranium dioxide pellets in Zircaloy cladding as the fuel material. Reactors that are fueled with uranium dioxide generally use water as the moderator. If ordinary water is used, the reactors are called Light Water Reactors (LWR), while if water enriched in the deuterium isotope of hydrogen is used, the reactors are called Heavy Water reactors. The LWRs can be either pressurized reactors (PWR) or boiling water reactors (BWR). Both of these reactor types use uranium that has been enriched in the 235 isotope to about 3.5 to 4% total abundance. There may be minor differences in the details of the spent fuel characteristics for PWRs and BWRs, but for simplicity we will not consider these second-order effects. The Canadian designed reactor (CANDU) that is moderated by heavy water uses natural uranium without enrichment of the 235 isotope as the fuel. These reactors run at higher linear power density than LWRs and produce spent fuel with lower total burn-up than LWRs. Where these difference are important with respect to spent fuel management, we will discuss them. Otherwise, we will concentrate on spent fuel from LWRs.

Oversby, V.M.

1996-06-01T23:59:59.000Z

363

Nuclear reactor control room construction  

DOE Patents (OSTI)

A control room for a nuclear plant is disclosed. In the control room, objects labelled 12, 20, 22, 26, 30 in the drawing are no less than four inches from walls labelled 10.2. A ceiling contains cooling fins that extend downwards toward the floor from metal plates. A concrete slab is poured over the plates. Studs are welded to the plates and are encased in the concrete. 6 figures.

Lamuro, R.C.; Orr, R.

1993-11-16T23:59:59.000Z

364

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

Dickson, J.J.

1963-09-24T23:59:59.000Z

365

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

Bassett, C.H.

1961-11-21T23:59:59.000Z

366

Behavior of spent nuclear fuel and storage system components in dry interim storage. Revision 1  

SciTech Connect

Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom; organic-cooled reactor (OCR) fuel (clad with a zirconium alloy) in silos in Canada; and boiling water reactor (BWR) fuel (clad with Zircaloy) in a metal storage cask in Germany. Dry storage demonstrations are under way for Zircaloy-clad fuel from BWRs, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions. 110 refs., 22 figs., 28 tabs.

Johnson, A.B. Jr.; Gilbert, E.R.; Guenther, R.J.

1983-02-01T23:59:59.000Z

367

CRC handbook of nuclear reactors calculations. Vol. II  

Science Conference Proceedings (OSTI)

This handbook breaks down the complex field of nuclear reactor calculations into major steps. Each step presents a detailed analysis of the problems to be solved, the parameters involved, and the elaborate computer programs developed to perform the calculations. This book bridges the gap between nuclear reactor theory and the implementation of that theory, including the problems to be encountered and the level of confidence that should be given to the methods described. Volume II: Monte Carlo Calculations for Nuclear Reactors. In-Core Management of Four Reactor Types. In-Core Management in CANDU-PHW Reactors. Reactor Dynamics. The Theory of Neutron Leakage in Reactor Lattices. Index.

Ronen, Y.

1986-01-01T23:59:59.000Z

368

Rodded shutdown system for a nuclear reactor  

DOE Patents (OSTI)

A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core.

Golden, Martin P. (Penn Township, Allegheny County, PA); Govi, Aldo R. (Greensburg, PA)

1978-01-01T23:59:59.000Z

369

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

Science Conference Proceedings (OSTI)

The U.S. Department of Energy decisions for the ultimate disposition of its inventory of used nuclear fuel presently in, and to be received and stored in, the L Basin at the Savannah River Site, and schedule for project execution have not been established. A logical decision timeframe for the DOE is following the review of the overall options for fuel management and disposition by the Blue Ribbon Commission on America's Nuclear Future (BRC). The focus of the BRC review is commercial fuel; however, the BRC has included the DOE fuel inventory in their review. Even though the final report by the BRC to the U.S. Department of Energy is expected in January 2012, no timetable has been established for decisions by the U.S. Department of Energy on alternatives selection. Furthermore, with the imminent lay-up and potential closure of H-canyon, no ready path for fuel disposition would be available, and new technologies and/or facilities would need to be established. The fuel inventory in wet storage in the 3.375 million gallon L Basin is primarily aluminum-clad, aluminum-based fuel of the Materials Test Reactor equivalent design. An inventory of non-aluminum-clad fuel of various designs is also stored in L Basin. Safe storage of fuel in wet storage mandates several high-level 'safety functions' that would be provided by the Structures, Systems, and Components (SSCs) of the storage system. A large inventory of aluminum-clad, aluminum-based spent nuclear fuel, and other nonaluminum fuel owned by the U.S. Department of Energy is in wet storage in L Basin at the Savannah River Site. An evaluation of the present condition of the fuel, and the Structures, Systems, or Components (SSCs) necessary for its wet storage, and the present programs and storage practices for fuel management have been performed. Activities necessary to validate the technical bases for, and verify the condition of the fuel and the SSCs under long-term wet storage have also been identified. The overall conclusion is that the fuel can be stored in L Basin, meeting general safety functions for fuel storage, for an additional 50 years and possibly beyond contingent upon continuation of existing fuel management activities and several augmented program activities. It is concluded that the technical bases and well-founded technologies have been established to store spent nuclear fuel in the L Basin. Methodologies to evaluate the fuel condition and characteristics, and systems to prepare fuel, isolate damaged fuel, and maintain water quality storage conditions have been established. Basin structural analyses have been performed against present NPH criteria. The aluminum fuel storage experience to date, supported by the understanding of the effects of environmental variables on materials performance, demonstrates that storage systems that minimize degradation and provide full retrievability of the fuel up to and greater than 50 additional years will require maintaining the present management programs, and with the recommended augmented/additional activities in this report.

Sindelar, R.; Deible, R.

2011-04-27T23:59:59.000Z

370

Activities of the OECD/NEA Expert Group on Assay Data for Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

Management of spent nuclear fuel is a key issue for many NEA member countries. In nuclear criticality safety, the decision of many countries to advance burnup credit as part of their licensing strategy has heightened recent interest in experimental data needed to validate computer codes used in burnup credit calculations. This paper discusses recent activities of an Expert Group on assay data, formed under the OECD/NEA/NSC/WPNCS (Working Party on Nuclear Criticality Safety) to help coordinate isotopic assay data activities and facilitate international collaboration between NEA member countries developing or implementing burnup credit methodologies. Recent activities of the Expert Group are described, focusing on the planned expansion of the Spent Fuel Isotopic Composition Database (SFCOMPO), and preparation of a state-of-the-art report on assay data that includes sections on recommended radiochemical analysis methods, techniques, and lessons learned from previous experiments.

Gauld, Ian C [ORNL; Rugama, Yolanda [OECD Nuclear Energy Agency

2009-01-01T23:59:59.000Z

371

Pyrochemical recovery of actinide elements from spent light water reactor fuel  

Science Conference Proceedings (OSTI)

Argonne National Laboratory is investigating salt transport and lithium pyrochemical processes for recovery of transuranic (TRU) elements from spent light water reactor fuel. The two processes are designed to recover the TRU elements in a form compatible with the Integral Fast Reactor (IFR) fuel cycle. The IFR is uniquely effective in consuming these long-lived TRU elements. The salt transport process uses calcium dissolved in Cu-35 wt % Mg in the presence of a CaCl{sub 2} salt to reduce the oxide fuel. The reduced TRU elements are separated from uranium and most of the fission products by using a MgCl{sub 2} transport salt. The lithium process, which does not employ a solvent metal, uses lithium in the presence of a LiCl salt as the reductant. After separation from the salt, the reduced metal is introduced into an electrorefiner, which separates the TRU elements from the uranium and fission products. In both processes, reductant and reduction salt are recovered by electrochemical decomposition of the oxide reaction product.

Johnson, G.K.; Pierce, R.D.; Poa, D.S.; McPheeters, C.C.

1994-01-01T23:59:59.000Z

372

Heavy Liquid Metal Reactor Development - Nuclear Engineering Division  

NLE Websites -- All DOE Office Websites (Extended Search)

> Heavy Liquid Metal Reactor Development > Heavy Liquid Metal Reactor Development Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Nuclear Data Program Advanced Reactor Development Overview Advanced Fast Reactor (AFR) Heavy Liquid Metal Reactor Development Generation IV Nuclear Waste Form and Repository Performance Modeling Nuclear Energy Systems Design and Development Other Capabilities Work with Argonne Contact us For Employees Site Map Help Advanced Reactor Development and Technology Heavy Liquid Metal Reactor Development Bookmark and Share STAR-LM: Simplified, Modular, Small Reactor Featuring Flow-thru Fuel Cartridge STAR-LM: Simplified, Modular, Small Reactor Featuring Flow-thru Fuel Cartridge. Click on image to view larger image. Argonne has traditionally been the foremost institute in the US for

373

Accelerator Laboratory AGN-201M Nuclear Reactor Laboratory  

E-Print Network (OSTI)

Laboratory Nuclear Power Institute (NPI) Nuclear Science Center (1MW Triga Reactor) (NSC) Nuclear SecurityAccelerator Laboratory AGN-201M Nuclear Reactor Laboratory Center for Large-scale Scientific Simulations (CLASS) Fuel Cycle and Materials Laboratory (FCML) Institute for National Security, Education

374

Radiation Shielding Analysis for Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors (DUPIC)  

Science Conference Proceedings (OSTI)

As a part of the compatibility analysis of DUPIC fuel in Canada deuterium uranium (CANDU) reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, which was originally designed for natural uranium core. At first, the conventional CANDU primary shield analysis method was validated using the Monte Carlo code MCNP-4B in order to assess the current analysis code system and the cross-section data. The computational benchmark calculation was performed for the CANDU end shield system, which has shown that the conventional method produces results consistent with the reference calculations as far as the total dose rate and total heat deposition rate are concerned. Second, the primary shield system analysis was performed for the DUPIC fuel core based on the power distribution obtained from the time-average core model, and the results have shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of the natural uranium core because the power levels on the core periphery are similar for both cores. This study has shown that the current primary shield system is adaptable for the DUPIC fuel CANDU core without design modification.

Roh, Gyuhong; Choi, Hangbok [Korea Atomic Energy Research Institute (Korea, Republic of)

2004-06-15T23:59:59.000Z

375

The Changing Landscape for Management of Spent Nuclear Fuel: International Perspectives from the OECD/NEA  

E-Print Network (OSTI)

important evolutions in the nuclear energy and waste management arenas. As we prepare to explore these topics in the coming days, it is useful to remind ourselves of the fundamental issues we face, and to consider the conclusions in 2006 and the major changes in context and perspectives since that time. Why are we concerned about spent nuclear fuel? The importance of safe and sustainable management of spent nuclear fuel is evident. While it comprises only a small amount by volume of the waste from nuclear power plants, it contains most of the radioactivity in national waste inventories. Its properties mean that special management is needed both in the near term as well as far into the future. The challenges are growing as greater volumes of SNF are foreseen to be stored for longer periods of time. Furthermore, SNF is at the heart of debates over nuclear power. At the last conference, nuclear power appeared poised to make a resurgence world-wide in response to, among other factors, desires for greater energy security and concerns over global warming. These factors have become even more prominent over the intervening years. Nuclear power is being expanded and extended in countries where it already exists. In addition, “newcomer ” states seeking sustainable and secure energy solutions are pursuing nuclear power.

Uichiro Yoshimura

2010-01-01T23:59:59.000Z

376

Multiple microprocessor based nuclear reactor power monitor  

SciTech Connect

The reactor power monitor is a portable multiple-microprocessor controlled data acquisition device being built for the International Atomic Energy Association. Its function is to measure and record the hourly integrated operating thermal power level of a nuclear reactor for the purpose of detecting unannounced plutonium production. The monitor consists of a /sup 3/He proportional neutron detector, a write-only cassette tape drive and control electronics based on two INTEL 8748 microprocessors. The reactor power monitor operates from house power supplied by the plant operator, but has eight hours of battery backup to cover power interruptions. Both the hourly power levels and any line power interruptions are recorded on tape and in memory. Intermediate dumps from the memory to a data terminal or strip chart recorder can be performed without interrupting data collection.

Lewis, P.S.; Ethridge, C.D.

1979-01-01T23:59:59.000Z

377

DOE fundamentals handbook: Nuclear physics and reactor theory. Volume 1  

Science Conference Proceedings (OSTI)

The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance.

Not Available

1993-01-01T23:59:59.000Z

378

DOE fundamentals handbook: Nuclear physics and reactor theory  

SciTech Connect

The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance.

Not Available

1993-01-01T23:59:59.000Z

379

DOE fundamentals handbook: Nuclear physics and reactor theory. Volume 2  

SciTech Connect

The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance.

Not Available

1993-01-01T23:59:59.000Z

380

FUEL ELEMENT FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

A fuel element structure particularly useful in high temperature nuclear reactors is presented. Basically, the structure comprises two coaxial graphite sleeves integrally joined together by radial fins. Due to the high structural strength of graphite at high temperatures and the rigidity of this structure, nuclear fuel encased within the inner sleeve in contiguous relation therewith is supported and prevented from expanding radially at high temperatures. Thus, the necessity of relying on the usual cladding materials with relatively low temperature limitations for structural strength is removed. (AEC)

Davidson, J.K.

1963-11-19T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR  

DOE Patents (OSTI)

The patent relates to the direct conversion of fission heat to electricity by use of thermionic plasma diodes having fissionable material cathodes, said diodes arranged to form a critical mass in a nuclear reactor. The patent describes a fuel element comprising a plurality of diodes each having a fissionable material cathode, an anode around said cathode, and an ionizable gas therebetween. Provision is made for flowing the gas and current serially through the diodes. (AEC)

Rasor, N.S.; Hirsch, R.L.

1963-12-01T23:59:59.000Z

382

Modeled Salt Density for Nuclear Material Estimation in the Treatment of Spent Nuclear Fuel  

SciTech Connect

Spent metallic nuclear fuel is being treated in a pyrometallurgical process that includes electrorefining the uranium metal in molten eutectic LiCl-KCl as the supporting electrolyte. We report a model for determining the density of the molten salt. Inventory operations account for the net mass of salt and for the mass of actinides present. It was necessary to know the molten salt density but difficult to measure, and it was decided to model the salt density for the initial treatment operations. The model assumes that volumes are additive for the ideal molten salt solution as a starting point; subsequently a correction factor for the lanthanides and actinides was developed. After applying the correction factor, the percent difference between the net salt mass in the electrorefiner and the resulting modeled salt mass decreased from more than 4.0% to approximately 0.1%. As a result, there is no need to measure the salt density at 500 C for inventory operations; the model for the salt density is found to be accurate.

DeeEarl Vaden; Robert. D. Mariani

2010-09-01T23:59:59.000Z

383

Environmental Statements, Availability, Etc., Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

8679 8679 Thursday June 1, 1995 Part III Department of Energy Environmental Statements, Availability, Etc.; Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs: Notice 28680 Federal Register / Vol. 60, No. 105 / Thursday, June 1, 1995 / Notices DEPARTMENT OF ENERGY Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs AGENCY: Department of Energy. ACTION: Record of decision. SUMMARY: The Department of Energy has issued a Record of Decision on Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs. The Record of Decision includes a Department-wide decision to

384

FIELD-DEPLOYABLE SAMPLING TOOLS FOR SPENT NUCLEAR FUEL INTERROGATION IN LIQUID STORAGE  

SciTech Connect

Methodology and field deployable tools (test kits) to analyze the chemical and microbiological condition of aqueous spent fuel storage basins and determine the oxide thickness on the spent fuel basin materials were developed to assess the corrosion potential of a basin. this assessment can then be used to determine the amount of time fuel has spent in a storage basin to ascertain if the operation of the reactor and storage basin is consistent with safeguard declarations or expectations and assist in evaluating general storage basin operations. The test kit was developed based on the identification of key physical, chemical and microbiological parameters identified using a review of the scientific and basin operations literature. The parameters were used to design bench scale test cells for additional corrosion analyses, and then tools were purchased to analyze the key parameters. The tools were used to characterize an active spent fuel basin, the Savannah River Site (SRS) L-Area basin. The sampling kit consisted of a total organic carbon analyzer, an YSI multiprobe, and a thickness probe. The tools were field tested to determine their ease of use, reliability, and determine the quality of data that each tool could provide. Characterization confirmed that the L Area basin is a well operated facility with low corrosion potential.

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

2012-09-12T23:59:59.000Z

385

CRC handbook of nuclear reactors calculations. Vol. III  

Science Conference Proceedings (OSTI)

This handbook breaks down the complex field of nuclear reactor calculations into major steps. Each step presents a detailed analysis of the problems to be solved, the parameters involved, and the elaborate computer programs developed to perform the calculations. This book bridges the gap between nuclear reactor theory and the implementation of that theory, including the problems to be encountered and the level of confidence that should be given to the methods described. Volume III: Control Rods and Burnable Absorber Calculations. Perturbation Theory for Nuclear Reactor Analysis. Thermal Reactors Calculations. Fast Reactor Calculations. Seed-Blanket Reactors. Index.

Ronen, Y.

1986-01-01T23:59:59.000Z

386

Advances in Development of the Fission Product Extraction Process for the Separation of Cesium and Strontium from Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

The Fission Product Extraction (FPEX) Process is being developed as part of the United States Department of Energy Advanced Fuel Cycle Initiative for the simultaneous separation of cesium (Cs) and strontium (Sr) from spent light water reactor (LWR) fuel. Separation of the Cs and Sr will reduce the short-term heat load in a geological repository, and when combined with the separation of americium (Am) and curium (Cm), could increase the capacity of the geological repository by a factor of approximately 100. The FPEX process is based on two highly specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. Results of flowsheet testing of the FPEX process with a simulated feed solution in 3.3-cm centrifugal contactors are detailed. Removal efficiencies, distribution coefficient data, coextraction of metals, and process hydrodynamic performance are discussed along with recommendations for future flowsheet testing with actual spent nuclear fuel.

JAck D. Law

2007-09-01T23:59:59.000Z

387

Melted and Granulated Depleted Uranium Dioxide for Use in Containers for Spent Nuclear Fuel  

NLE Websites -- All DOE Office Websites (Extended Search)

Melted and Granulated Depleted Uranium Dioxide for Use in Containers for Spent Nuclear Fuel Melted and Granulated Depleted Uranium Dioxide for Use in Containers for Spent Nuclear Fuel Vitaly T. Gotovchikov a , Victor A. Seredenko a , Valentin V. Shatalov a , Vladimir N. Kaplenkov a , Alexander S. Shulgin a , Vladimir K. Saranchin a , Michail A. Borik a∗ , Charles W. Forsberg b , All-Russian Research Institute of Chemical Technology (ARRICT) 33, Kashirskoe ave., Moscow, Russia, 115409, E-mail: chem.conv@ru.net Oak Ridge National Laboratory (ORNL) Bethel Wall Road, P.O. Box 2008, MS-6165, Oak Ridge, TN, USA, 37831 Abstract - Induction cold crucible melters (ICCM) have the potential to be a very-low-cost high-throughput method for the production of DUO 2 for SNF casks. The proposed work would develop these melters for this specific application. If a

388

Plan for characterization of K Basin spent nuclear fuel and sludge  

SciTech Connect

This plan outlines a characterization program that supports the accelerated Path Forward scope and schedules for the Spent Nuclear Fuel stored in the Hanford K Basins. This plan is driven by the schedule to begin fuel transfer by December 1997. The program is structured for 4 years and is limited to in-situ and laboratory examinations of the spent nuclear fuel and sludge in the K East and K West Basins. The program provides bounding behavior of the fuel, and verification and acceptability for three different sludge disposal pathways. Fuel examinations are based on two shipping campaigns for the K West Basin and one from the K East Basin. Laboratory examinations include physical condition, hydride and oxide content, conditioning testing, and dry storage behavior.

Lawrence, L.A.; Marschman, S.C.

1995-06-01T23:59:59.000Z

389

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

DOE Patents (OSTI)

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.

Heckman, R.A.

1980-12-19T23:59:59.000Z

390

Security of Nuclear Reactors and Special Nuclear Materials This revisiono  

E-Print Network (OSTI)

Provides requirements for the recovery of lost, seized, or stolen special nuclear material (para 2-1b). o Prescribes that unclassified information pertaining to plans, procedures, and equipment for the physical protection of nuclear reactors and special nuclear material will be safeguarded as DoD Unclassified Controlled Nuclear Information (para 2-1f). o Requires the conduct of a vulnerability assessment at each facility where special nuclear material is used or stored (para 2-2a). o Provides that Headquarters, U. S. Army Materiel Command will develop the postulated threat as the basis for the vulnerability assessment (para 2-2b), as well as the standardized format for documenting the results of the assessment and for the after action reports (para 2-2h). o Designates special nuclear material as inherently dangerous to others for use of force purposes (para 2-4a). o Prescribes minimum storage standards for special nuclear material (para 3-1). o Provides for the protection of vital equipment (para 3-3). o Explains the concept of the required security system for nuclear reactors and special nuclear material (para 4-2). o Establishes specific physical security standards for the protection of nuclear reactors and special nuclear material (para 4-4), to include required access controls (para 4-5). o Prohibits the locksmith from being designated as the key control officer or lock custodian (para 4-5g(25)). o Provides guidance on meeting requirement to continuously man two alarm monitoring facilities (para 4-6b). o Allows continued use of monitoring console systems installed prior to publication of this regulation that do not meet the map or video display requirement (para 4-6g(1)). o Provides guidance for testing the perimeter intrusion detection system (para 4-6n(2)). o Requires appropriate security personnel be trained to manually start the standby generator if the automatic starter fails to function properly (para 4-9b(4)). o Provides that the size, composition, and response time of the response force will be set by the major subordinate commander and approved by the Commanding

unknown authors

1993-01-01T23:59:59.000Z

391

Separation of uranium from technetium in recovery of spent nuclear fuel  

DOE Patents (OSTI)

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

Pruett, David J. (Knoxville, TN); McTaggart, Donald R. (Knoxville, TN)

1984-01-01T23:59:59.000Z

392

Handbook of Neutron Absorber Materials for Spent Nuclear Fuel Transportation and Storage Applications  

Science Conference Proceedings (OSTI)

This handbook is intended to become a single source of information regarding technical characteristics of neutron absorber materials that have been used for storage and transportation of spent nuclear fuel as well as to provide a summary of users' experience. The second edition of this handbook was published in 2006. This third edition, the 2009 Edition, updates materials covered in the 2006 Edition, presents new products introduced since 2006, and reflects recent realignments of neutron absorber suppliers.

2009-11-25T23:59:59.000Z

393

Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes  

SciTech Connect

A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs.

Forsberg, C.W.

1997-03-01T23:59:59.000Z

394

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

SciTech Connect

The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. Because this sub-project is still in the construction/start-up phase, all verification activities have not yet been performed (e.g., canister cover cap and welding fixture system verification, MCO Internal Gas Sampling equipment verification, and As-built verification.). The verification activities identified in this report that still are to be performed will be added to the start-up punchlist and tracked to closure.

PICKETT, W.W.

2000-09-22T23:59:59.000Z

395

Disposition Options for Hanford Site K-Basin Spent Nuclear Fuel Sludge  

SciTech Connect

This report provides summary-level information about a group of options that have been identified for the disposition of spent-nuclear-fuel sludge in the K-Basins at the Hanford Site. These options are representative of the range of likely candidates that may be considered for disposition of the sludge. The product of each treatment option would be treated sludge that would meet waste acceptance requirements for disposal as transuranic (TRU) waste at the Waste Isolation Pilot Plant (WIPP).

Mellinger, George B.; Delegard, Calvin H.; Gerber, Mark A.; Naft, Barry N.; Schmidt, Andrew J.; Walton, Terry L.

2004-01-18T23:59:59.000Z

396

Separation of uranium from technetium in recovery of spent nuclear fuel  

DOE Patents (OSTI)

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

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

1983-08-31T23:59:59.000Z

397

Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Safety Culture in the US Nuclear Regulatory Commission's Reactor Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process September 19, 2012 Presenter: Undine Shoop, Chief, Health Physics and Human Performance Branch, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission Topics covered: Purpose of the Reactor Oversight Process (ROP) ROP Framework Safety Culture within the ROP Safety Culture Assessments Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process More Documents & Publications A Commissioner's Perspective on USNRC Actions in Response to the Fukushima Nuclear Accident Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities, 2/17/11

398

Advanced nuclear reactor public opinion project  

SciTech Connect

This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

Benson, B.

1991-07-25T23:59:59.000Z

399

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network (OSTI)

Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in electron anti-neutrino detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties and their relevance to reactor anti-neutrino experiments.

Z. Djurcic; J. A. Detwiler; A. Piepke; V. R. Foster Jr.; L. Miller; G. Gratta

2008-08-06T23:59:59.000Z

400

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network (OSTI)

Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in electron anti-neutrino detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties and their relevance to reactor anti-neutrino experiments.

Djurcic, Z; Piepke, A; Foster, V R; Miller, L; Gratta, G

2008-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

Nuclear reactor pressure vessel support system  

DOE Patents (OSTI)

A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

Sepelak, George R. (McMurray, PA)

1978-01-01T23:59:59.000Z

402

Assessment of Incident-Free Transport for Transport of Spent Nuclear Fuel to Yucca Mountain Using RADTRAN 5.5  

Science Conference Proceedings (OSTI)

This report evaluates the incident-free radiological impacts associated with the transportation of spent nuclear fuel to the proposed Yucca Mountain repository using the RADTRAN 5.5 computer code developed by Sandia National Laboratories.

2005-09-28T23:59:59.000Z

403

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

SciTech Connect

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

D. Vinson; A. Serika

2002-08-26T23:59:59.000Z

404

What are Spent Nuclear Fuel and High-Level Radioactive Waste ?  

Science Conference Proceedings (OSTI)

Spent nuclear fuel and high-level radioactive waste are materials from nuclear power plants and government defense programs. These materials contain highly radioactive elements, such as cesium, strontium, technetium, and neptunium. Some of these elements will remain radioactive for a few years, while others will be radioactive for millions of years. Exposure to such radioactive materials can cause human health problems. Scientists worldwide agree that the safest way to manage these materials is to dispose of them deep underground in what is called a geologic repository.

DOE

2002-12-01T23:59:59.000Z

405

Spatial multi-taper spectrum estimation for nuclear reactor modelling  

Science Conference Proceedings (OSTI)

Multi-taper univariate and cross-spectral analysis is used to investigate the structure of spatial variation in the temperatures measured across the surface of a nuclear reactor. The construction of the spatial tapers over the approximate circular reactor ...

C. J. Scarrott; G. Tunnicliffe Wilson

2009-10-01T23:59:59.000Z

406

Spent nuclear fuel project multi-canister overpack, additional NRC requirements  

Science Conference Proceedings (OSTI)

The US Department of Energy (DOE), established in the K Basin Spent Nuclear Fuel Project Regulatory Policy, dated August 4, 1995 (hereafter referred to as the Policy), the requirement for new Spent Nuclear Fuel (SNF) Project facilities to achieve nuclear safety equivalency to comparable US Nuclear Regulatory Commission (NRC)-licensed facilities. For activities other than during transport, when the Multi-Canister Overpack (MCO) is used and resides in the Canister Storage Building (CSB), Cold Vacuum Drying (CVD) facility or Hot Conditioning System, additional NRC requirements will also apply to the MCO based on the safety functions it performs and its interfaces with the SNF Project facilities. An evaluation was performed in consideration of the MCO safety functions to identify any additional NRC requirements needed, in combination with the existing and applicable DOE requirements, to establish nuclear safety equivalency for the MCO. The background, basic safety issues and general comparison of NRC and DOE requirements for the SNF Project are presented in WHC-SD-SNF-DB-002.

Garvin, L.J.

1998-08-04T23:59:59.000Z

407

Fuel handling system for a nuclear reactor  

DOE Patents (OSTI)

A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

Saiveau, James G. (Hickory Hills, IL); Kann, William J. (Park Ridge, IL); Burelbach, James P. (Glen Ellyn, IL)

1986-01-01T23:59:59.000Z

408

Closure head for a nuclear reactor  

DOE Patents (OSTI)

A closure head for a nuclear reactor includes a stationary outer ring integral with the reactor vessel with a first rotatable plug disposed within the stationary outer ring and supported from the stationary outer ring by a bearing assembly. A sealing system is associated with the bearing assembly to seal the annulus defined between the first rotatable plug and the stationary outer ring. The sealing system comprises tubular seal elements disposed in the annulus with load springs contacting the tubular seal elements so as to force the tubular seal elements against the annulus in a manner to seal the annulus. The sealing system also comprises a sealing fluid which is pumped through the annulus and over the tubular seal elements causing the load springs to compress thereby reducing the friction between the tubular seal elements and the rotatable components while maintaining a gas-tight seal therebetween.

Wade, Elman E. (South Huntingdon, PA)

1980-01-01T23:59:59.000Z

409

Nuclear reactor insulation and preheat system  

DOE Patents (OSTI)

An insulation and preheat system for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the compartment. An external surface of the compartment or enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair.

Wampole, Nevin C. (Latrobe, PA)

1978-01-01T23:59:59.000Z

410

Nuclear reactor flow control method and apparatus  

DOE Patents (OSTI)

Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

Church, J.P.

1993-03-30T23:59:59.000Z

411

Neutrino Oscillation Experiments at Nuclear Reactors  

E-Print Network (OSTI)

In this paper I give an overview of the status of neutrino oscillation experiments performed using nuclear reactors as sources of neutrinos. I review the present generation of experiments (Chooz and Palo Verde) with baselines of about 1 km as well as the next generation that will search for oscillations with a baseline of about 100 km. While the present detectors provide essential input towards the understanding of the atmospheric neutrino anomaly, in the future, the KamLAND reactor experiment represents our best opportunity to study very small mass neutrino mixing in laboratory conditions. In addition KamLAND with its very large fiducial mass and low energy threshold, will also be sensitive to a broad range of different physics.

Giorgio Gratta

1999-05-06T23:59:59.000Z

412

Nuclear reactor flow control method and apparatus  

DOE Patents (OSTI)

This document describes method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

Church, J.P.

1991-04-23T23:59:59.000Z

413

Reactor Physics and Fuel Cycle Analysis - Nuclear Engineering Division  

NLE Websites -- All DOE Office Websites (Extended Search)

Analysis Analysis Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Overview Current Projects Software Nuclear Plant Dynamics and Safety Nuclear Data Program Advanced Reactor Development Nuclear Waste Form and Repository Performance Modeling Nuclear Energy Systems Design and Development Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Reactor Physics and Fuel Cycle Analysis Bookmark and Share Reactor physics and fuel cycle analysis is a core competency of the Nuclear Engineering (NE) Division. The Division has played a major role in the design and analysis of advanced reactors, particularly liquid-metal-cooled reactors. NE researchers have concentrated on developing computer codes for

414

Nuclear Thermal Rockets: The Physics of the Fission Reactor  

E-Print Network (OSTI)

Nuclear Thermal Rockets: The Physics of the Fission Reactor Shane D. Ross Control and Dynamical combustion are those powered by nuclear fission. Comparison of Chemical and Nuclear Rockets. Most existent.g., hydrogen and oxygen). In a nuclear rocket, or more precisely, a nuclear thermal rocket, the propellant

Ross, Shane

415

Truck and rail charges for shipping spent fuel and nuclear waste  

SciTech Connect

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

McNair, G.W.; Cole, B.M.; Cross, R.E.; Votaw, E.F.

1986-06-01T23:59:59.000Z

416

Ground test facility for nuclear testing of space reactor subsystems  

SciTech Connect

Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs.

Quapp, W.J.; Watts, K.D.

1985-01-01T23:59:59.000Z

417

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel  

E-Print Network (OSTI)

and Nuclear Recoil . . . . . . . . . . . . . . . . . . . . .2 Quantitative Measurements using NRF 2.1 Nuclear ResonanceFuture Work A Transmission Nuclear Resonance Fluorescence

Ludewigt, Bernhard A

2011-01-01T23:59:59.000Z

418

Measuring Neutrino Oscillations with Nuclear Reactors  

SciTech Connect

Since the first direct observations of antineutrino events by Reines and Cowan in the 1950's, nuclear reactors have been an important tool in the study of neutrino properties. More recently, the study of neutrino oscillations has been a very active area of research. The pioneering observation of oscillations by the KamLAND experiment has provided crucial information on the neutrino mixing matrix. New experiments to study the remaining unknown mixing angle are currently under development. These recent studies and potential future developments will be discussed.

McKeown, R. D. [W. K. Kellogg Radiation Laboratory, California Institute of Technology, Pasadena, CA (United States)

2007-10-26T23:59:59.000Z

419

Nuclear reactor containment spray testing system. [PWR  

SciTech Connect

Disclosed is a method for periodic testing of a spray system in a nuclear reactor containment. The method includes injecting a gas into the spray system such that a temperature differential exists between the gas and the containment atmosphere. Scanning the gas jet discharged from the spray nozzles with infrared apparatus then provides a real-time thermal image on a monitor, such as a cathode ray tube, and detects any partially or completely blocked nozzles in the spray system. The scanning may be performed from the containment operating deck. 1 claim, 4 figures.

Rubin, K.

1978-01-10T23:59:59.000Z

420

Liquid metal pump for nuclear reactors  

DOE Patents (OSTI)

A pump for use in pumping high temperature liquids at high pressures, particularly liquid metals used to cool nuclear reactors is described. It is of the type in which the rotor is submerged in a sump but is fed by an inlet duct which bypasses the sump. A chamber, kept full of fluid, surrounds the pump casing into which fluid is bled from the pump discharge and from which fluid is fed to the rotor bearings and hence to the sump. This equalizes pressure inside and outside the pump casing and reduces or eliminates the thermal shock to the bearings and sump tank.

Allen, H.G.; Maloney, J.R.

1975-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor spent nuclear" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

Dry halide method for separating the components of spent nuclear fuels  

DOE Patents (OSTI)

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

Christian, Jerry Dale (Idaho Falls, ID); Thomas, Thomas Russell (Rigby, ID); Kessinger, Glen F. (Idaho Falls, ID)

1998-01-01T23:59:59.000Z

422

Simulated Performance of the Integrated Passive Neutron Albedo Reactivity and Self-Interrogation Neutron Resonance Densitometry Detector Designed for Spent Fuel Measurement at the Fugen Reactor in Japan  

Science Conference Proceedings (OSTI)

An integrated nondestructive assay instrument, which combined the Passive Neutron Albedo Reactivity (PNAR) and the Self-Interrogation Neutron Resonance Densitometry (SINRD) techniques, is the research focus for a collaborative effort between Los Alamos National Laboratory (LANL) and the Japanese Atomic Energy Agency as part of the Next Generation Safeguard Initiative. We will quantify the anticipated performance of this experimental system in two physical environments: (1) At LANL we will measure fresh Low Enriched Uranium (LEU) assemblies for which the average enrichment can be varied from 0.2% to 3.2% and for which Gd laced rods will be included. (2) At Fugen we will measure spent Mixed Oxide (MOX-B) and LEU spent fuel assemblies from the heavy water moderated Fugen reactor. The MOX-B assemblies will vary in burnup from {approx}3 GWd/tHM to {approx}20 GWd/tHM while the LEU assemblies ({approx}1.9% initial enrichment) will vary from {approx}2 GWd/tHM to {approx}7 GWd/tHM. The estimated count rates will be calculated using MCNPX. These preliminary results will help the finalization of the hardware design and also serve a guide for the experiment. The hardware of the detector is expected to be fabricated in 2012 with measurements expected to take place in 2012 and 2013. This work is supported by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.

Ulrich, Timothy J. II [Los Alamos National Laboratory; Lafleur, Adrienne M. [Los Alamos National Laboratory; Menlove, Howard O. [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory; Seya, Michio [Los Alamos National Laboratory; Bolind, Alan M. [Los Alamos National Laboratory

2012-07-16T23:59:59.000Z

423

U.S. Spent Nuclear Fuel Data as of December 31,2002 -Table 2  

Gasoline and Diesel Fuel Update (EIA)

6 6 Table 1 | Table 3 Table 2. Nuclear Power Plant Data as of December 31, 2002 Reactor Name State Reactor Type Reactor Vendor a Core Size (number of assemblies) Startup Date (year)b License Expiration (year) Actual Retirement (year) Arkansas Nuclear 1 AR PWR B&W 177 1974 2034 Arkansas Nuclear 2 AR PWR CE 177 1978 2018 Beaver Valley 1 PA PWR WE 157 1976 2016 Beaver Valley 2 PA PWR WE 157 1987 2027 Big Rock Point MI BWR GE 84 1962 2000 1998 Braidwood 1 IL PWR WE 193 1987 2026 Braidwood 2 IL PWR WE 193 1988 2027 Browns Ferry 1 AL BWR GE 764 1973 2013 Browns Ferry 2 AL BWR GE 764 1974 2014 Browns Ferry 3 AL BWR GE 764 1976 2016 Brunswick 1 NC BWR GE 560 1976 2016

424

More About NNSA's Naval Reactors Office | National Nuclear Security...  

National Nuclear Security Administration (NNSA)

to skip to the main content Facebook Flickr RSS Twitter YouTube More About NNSA's Naval Reactors Office | National Nuclear Security Administration Our Mission Managing the...

425

TABLE 1. Nuclear Reactor, State, Type, Net Capacity ...  

U.S. Energy Information Administration (EIA)

Nuclear Reactor, State, Type, Net Capacity, ... Quad Cities Generating Station River Bend San Onofre Seabrook Sequoyah South Texas Project St Lucie ...

426

Characterization of Nuclear Reactor Materials and Components with ...  

Science Conference Proceedings (OSTI)

Mar 6, 2013 ... Characterization of Nuclear Reactor Materials and Components with ... Results are discussed in terms of existing theoretical models for hydride ...

427

Light Water Reactor Materials for Commercial Nuclear Power ...  

Science Conference Proceedings (OSTI)

Presentation Title, Light Water Reactor Materials for Commercial Nuclear ... First- Principles Theory of Magnetism, Crystal Field and Phonon Spectrum of UO2.

428

Nuclear Reactor Materials at the Atomic Scale - Programmaster.org  

Science Conference Proceedings (OSTI)

Presentation Title, Nuclear Reactor Materials at the Atomic Scale ... Study of the Interaction of Solutes with Interfaces in Iron Using Density-Functional Theory.

429

Fuel performance comparison between Savannah River reactors and the US commercial nuclear reactors  

SciTech Connect

This document provides a review of fuel/target performance of the Savannah River Reactors which was made to compare their in-core performance with that of the commercial nuclear reactors in the US.

Paik, I.K.; Ellison, P.G.

1989-01-01T23:59:59.000Z

430

Fluid sampling system for a nuclear reactor  

DOE Patents (OSTI)

A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump. 1 fig.

Lau, L.K.; Alper, N.I.

1994-11-22T23:59:59.000Z

431

PHYSICS OF NUCLEAR REACTORS Nuclear reactions and cross sections 1-10  

E-Print Network (OSTI)

PHYSICS OF NUCLEAR REACTORS Nuclear reactions and cross sections 1-10 10 11 12 13 14 15 16 17 18 19 neutron wavelength, D is given by: cE mM Mm 2 + = h D , (1.22) 1 Bell and Glasstone, Nuclear Reactor Theory, p. 392, 1970. #12;PHYSICS OF NUCLEAR REACTORS Nuclear reactions and cross sections 1-11 Where m

Danon, Yaron

432

Evaluation of Shortline Railroads & SNF/HLW Rail Shipment Inspections Tasked for the Transportation of Spent Nuclear Fuel  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Transportation Transportation Stakeholders National Transportation Stakeholders National Transportation Stakeholders National Transportation Stakeholders Forum Forum 2011 Annual Meeting 2011 Annual Meeting 2011 Annual Meeting 2011 Annual Meeting May 11, 2011 May 11, 2011 Evaluation of Shortline Railroads Evaluation of Shortline Railroads & & & & SNF/HLW Rail Shipment Inspections SNF/HLW Rail Shipment Inspections Tasked for the Transportation of Spent Nuclear Fuel Tasked for the Transportation of Spent Nuclear Fuel Evaluation of Shortline Railroads Evaluation of Shortline Railroads Evaluation of Shortline Railroads Evaluation of Shortline Railroads Task: Task: Task: Task: Identify Shortline Railroads Serving Nuclear Power Plants Identify Shortline Railroads Serving Nuclear Power Plants

433

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

DOE Patents (OSTI)

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

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

1982-01-19T23:59:59.000Z

434

On the Development of a Distillation Process for the Electrometallurgical Treatment of Irradiated Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

As part of the spent fuel treatment program at the Idaho National Laboratory, a vacuum distillation process is being employed for the recovery of actinide products following an electrorefining process. Separation of the actinide products from a molten salt electrolyte and cadmium is achieved by a batch operation called cathode processing. A cathode processor has been designed and developed to efficiently remove the process chemicals and consolidate the actinide products for further processing. This paper describes the fundamentals of cathode processing, the evolution of the equipment design, the operation and efficiency of the equipment, and recent developments at the cathode processor. In addition, challenges encountered during the processing of irradiated spent nuclear fuel in the cathode processor will be discussed.

B.R. Westphal; K.C. Marsden; J.C. Price; D.V. Laug

2008-04-01T23:59:59.000Z

435

Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL  

SciTech Connect

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

D. Kokkinos