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1

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

2

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

3

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

4

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

5

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

6

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

7

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

8

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

9

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

10

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 ...............................................................................................................................

11

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

12

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

13

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

14

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

15

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

16

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

17

Brookhaven Medical Research Reactor  

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

Medical Research Reactor BMRR The last of the Lab's reactors, the Brookhaven Medical Research Reactor (BMRR), was shut down in December 2000. The BMRR was a three megawatt...

18

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

19

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

20

Research reactors - an overview  

SciTech Connect

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

West, C.D.

1997-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" 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

Decommssioning Of Research Reactor: Problemsand Experience  

E-Print Network (OSTI)

The study of the preparation for decommissioning the Research Reactor in Salaspils (Latvia) and the experience of decommissioning the Research Reactor in Sosny (Belarus) show that the problem of decommissioning research reactors is acute for countries that have no NPPs or their own nuclear industry. It also is associated with regulatory framework, planning and design, dismantling technologies, decontamination of radioactive equipment and materials, spent fuel and radioactive waste management, etc. 1. INTRODUCTION According to the IAEA research reactor database, there are about 300 research reactors worldwide [1]. At present over 30% of them have lifetimes of more than 35 years, 60% of more than 25 years. After the Chernobyl accident, significant efforts were made by many countries to modernize old research reactors aiming, first of all, at ensuring safe operation. However, a large number of aging research reactors will be facing shutdown in the near future. The problem of decommis...

Alexander Mikhalevich

2000-01-01T23:59:59.000Z

22

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

23

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

24

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

25

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

26

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

27

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

28

Foreign Research Reactor/Domestic Research Reactor Receipt Coordinator...  

National Nuclear Security Administration (NNSA)

Research ReactorDomestic Research Reactor Receipt Coordinator, Savannah River Nuclear Solutions | National Nuclear Security Administration Our Mission Managing the Stockpile...

29

United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support  

SciTech Connect

The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

Douglas Morrell

2011-03-01T23:59:59.000Z

30

ADMINISTRATION OF ORNL RESEARCH REACTORS  

SciTech Connect

Organization of the ORNL Operations division for administration of the Oak Ridge Research Reactor, the Low Intensity Testing Reactor, and the Oak Ridge Graphite Reactor is described. (J.R.D.)

Casto, W.R.

1962-08-20T23:59:59.000Z

31

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

32

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

33

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

34

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

35

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

36

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

37

Advanced Reactor Research and Development Funding Opportunity...  

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

Advanced Reactor Research and Development Funding Opportunity Announcement Advanced Reactor Research and Development Funding Opportunity Announcement The U.S. Department of Energy...

38

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

39

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

40

2012 Annual Report Research Reactor Infrastructure Program  

SciTech Connect

The content of this report is the 2012 Annual Report for the Research Reactor Infrastructure Program.

Douglas Morrell

2012-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" 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

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):

42

International Research Reactor Decommissioning Project  

SciTech Connect

Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement to the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.

Leopando, Leonardo [Philippine Nuclear Research Institute, Quezon City (Philippines); Warnecke, Ernst [International Atomic Energy Agency, Vienna (Austria)

2008-01-15T23:59:59.000Z

43

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

44

Reactor sharing experience at the MIT research reactor  

SciTech Connect

This paper provides a number of examples of how educational institutions in the Boston area and elsewhere that do not possess nuclear reactors for training and research purposes have successfully enriched their programs through utilization of the Massachusetts Institute of Technology Research Reactor (MITR) with assistance from the Reactor Sharing Program of the US Department of Energy (DOE).

Clark, L. Jr.; Fecych, W.; Young, H.H.

1985-01-01T23:59:59.000Z

45

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

46

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

47

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

48

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

49

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

50

BNL | Our History: Reactors as Research Tools  

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

> See also: Accelerators > See also: Accelerators Brookhaven History: Using Reactors as Research Tools BGRR Brookhaven Graphite Research Reactor The Brookhaven Graphite Research Reactor (BGRR) was the Laboratory's first big machine and the first peace-time reactor built in the United States following World War II. The reactor's primary mission was to produce neutrons for scientific experimentation and to refine reactor technology. At the time, the BGRR could accommodate more simultaneous experiments than any other reactor. Scientists and engineers from every corner of the U.S. came to use the reactor, which was not only a source of neutrons for experiments, but also an excellent training facility. Researchers used the BGRR's neutrons as tools for studying atomic nuclei and the structure of solids, and to investigate many physical, chemical and

51

Research Reactors Division | Neutron Science | ORNL  

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

is responsible for operation of the High Flux Isotope Reactor. Operating at 85 MW, HFIR is the highest flux reactor-based source of neutrons for research in the United States,...

52

History of Research Reactors at Brookhaven  

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

History of Research Reactors at Brookhaven History of Research Reactors at Brookhaven Brookhaven National Laboratory has three nuclear reactors on its site that were used for scientific research. The reactors are all shut down, and the Laboratory is addressing environmental issues associated with their operations. photo of BGRR Brookhaven Graphite Research Reactor - Beginning operations in 1950, the graphite reactor was used for research in medicine, biology, chemistry, physics and nuclear engineering. One of the most significant achievements at this facility was the development of technetium-99m, a radiopharmaceutical widely used to image almost any organ in the body. The graphite reactor was shut down in 1969. Parts of it have been decommissioned, with the remainder to be addressed by 2011. More history

53

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

54

Brookhaven Graphite Research Reactor | Environmental Restoration...  

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

Brookhaven Graphite Research Reactor(BGRR) BGRR Overview BGRR Complex Description Decommissioning Decision BGRR Complex Cleanup Actions BGRR Documents BGRR Science &...

55

FOREIGN RESEARCH AND POWER REACTOR PRELIMINARY LIST  

SciTech Connect

Foreign research and power reactors are tabulated. Nuclear power buildup goals are given for each nation on which information is available. (J.H.D.)

Ullmann, J.W.

1959-02-26T23:59:59.000Z

56

Cleanup at the Brookhaven Graphite Research Reactor  

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

Graphite Research Reactor placeholder Remotely operated robot known as a BROKK manipulator In April 2005, the Department of Energy (DOE), the Lab, and the regulatory agencies...

57

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

58

University Research Reactor Task Force to the Nuclear Energy Research  

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

University Research Reactor Task Force to the Nuclear Energy University Research Reactor Task Force to the Nuclear Energy Research Advisory Committee University Research Reactor Task Force to the Nuclear Energy Research Advisory Committee In mid-February, 2001 The University Research Reactor (URR) Task Force (TF), a sub-group of the Department of Energy (DOE) Nuclear Energy Research Advisory Committee (NERAC), was asked to: * Analyze information collected by DOE, the NERAC "Blue Ribbon Panel," universities, and other sources pertaining to university reactors including their research and training capabilities, costs to operate, and operating data, and * Provide DOE with clear, near-term recommendations as to actions that should be taken by the Federal Government and a long-term strategy to assure the continued operation of vital university reactor facilities in

59

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

60

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

Note: This page contains sample records for the topic "research reactor spent" 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

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

62

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

63

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

64

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

65

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

66

Independent Verification of Research Reactor Operation (Analysis of the Georgian IRT-M Reactor by the Isotope Ratio Method)  

SciTech Connect

The U.S. Department of Energy’s Office of Nonproliferation and International Security (NA-24) develops technologies to aid in implementing international nuclear safeguards. The Isotope Ratio Method (IRM) was successfully developed in 2005 – 2007 by Pacific Northwest National Laboratory (PNNL) and the Republic of Georgia’s Andronikashvili Institute of Physics as a generic technology to verify the declared operation of water-moderated research reactors, independent of spent fuel inventory. IRM estimates the energy produced over the operating lifetime of a fission reactor by measuring the ratios of the isotopes of trace impurity elements in non-fuel reactor components.The Isotope Ratio Method is a technique for estimating the energy produced over the operating lifetime of a fission reactor by measuring the ratios of the isotopes of impurity elements in non-fuel reactor components.

Cliff, John B.; Frank, Douglas P.; Gerlach, David C.; Gesh, Christopher J.; Little, Winston W.; Reid, Bruce D.; Tsiklauri, Georgi V.; Abramidze, Sh; Rostomashvili, Z.; Kiknadze, G.; Dzhavakhishvily, O.; Nabakhtiani, G.

2010-08-11T23:59:59.000Z

67

PROCEEDINGS OF THE AEC SYMPOSIUM FOR CHEMICAL PROCESSING OF IRRADIATED FUELS FROM POWER, TEST, AND RESEARCH REACTORS, RICHLAND, WASHINGTON, OCTOBER 20 AND 21, 1959  

SciTech Connect

A review is presented in this symposium of the technology currently available for processing spent fuels from research, test, and power reactors. Twenty-one papers are included. Separate abstracts have been prepared for each paper. (W.L.H.)

1960-01-01T23:59:59.000Z

68

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

69

Relicensing of the MIT Research Reactor  

SciTech Connect

The Massachusetts Institute of Technology (MIT) Research Reactor (MITR) is owned and operated by MIT, a nonprofit university. The current reactor, MITR-II, is a 5-MW, light water-cooled and heavy water-moderated reactor that uses materials test reactor-type fuel. Documents supporting application to the U.S. Nuclear Regulatory Commission (NRC) for relicensing of MITR were submitted in July 1999. A power upgrade from 5 to 6 MW was also requested. The relicensed reactor (MITR-III) will be the third reactor operated by MIT. This paper describes MITR-I and MITR-II, and design options considered for MITR-III. Selected problems addressed during the relicensing studies are also described, namely core tank aging evaluation, neutronic analysis, thermal-hydraulic analysis, and step reactivity insertion analysis.

Lin-Wen Hu; John A. Bernard; Susan Tucker

2000-06-04T23:59:59.000Z

70

Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report  

SciTech Connect

This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed.

Hauptman, H.M.; Petro, J.N.; Jacobi, O. [and others

1995-04-01T23:59:59.000Z

71

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...

72

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

73

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

74

Brookhaven Graphite Research Reactor | Environmental Restoration Projects |  

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

Brookhaven Graphite Research Reactor Documents Brookhaven Graphite Research Reactor Documents Feasibility Study (PDF) Proposed Remedial Action Plan (PDF) Record of Decision (PDF) RD/RA Work Plan for the BGRR Pile (PDF) RD/RA Work Plan for the Bioshield (PDF) RD/RA Work Plan for the BGRR Cap (PDF) Brookhaven Graphite Research Reactor Explanation of Significant Differences (PDF) (4/12) NYSDEC Approval Letter for BGRR ESD (PDF) (5/12) USEPA Approval Letter for BGRR ESD (PDF) (6/12) DOE BGRR ESD Transmittal Letter (PDF) (7/12) Remedial Design Implementation Report (PDF) (12/11) Completion Reports Removal of the Above-Ground Ducts and Preparation of the Instrument House (708) for Removal (PDF) - April 2002 Below-Ground Duct Outlet Air Coolers, Filters and Primary Liner Removal (PDF) - April 2005 Canal and Deep Soil Pockets Excavation and Removal (PDF) - August

75

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

76

License extension for the MIT research reactor  

Science Conference Proceedings (OSTI)

On February 8, 1995, the U.S. Nuclear Regulatory Commission (NRC) amended the operating license for the Massachusetts Institute of Technology (MIT) research reactor (MITR) by extending its expiration date from May 7, 1996, to August 8, 1999. This increase of 2.25 yr in the duration of the license was for the purpose of recapturing construction time. Extensions of this type have routinely been granted to power plants. The significance of the present action is that it is the first such extension ever granted to a research reactor. This paper describes the basis for the amendment request and the supporting safety analysis.

Bernard, J.A. [Massachusetts Institute of Technology, Cambridge, MA (United States)

1995-12-31T23:59:59.000Z

77

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...

78

The triple axis spectrometer at the new research reactor OPAL ...  

Science Conference Proceedings (OSTI)

... The triple axis spectrometer at the new research reactor OPAL in Australia. ... The TAS will be based on a thermal beam at the reactor face. ...

79

Basic and Applied Science Research Reactors - Reactors designed...  

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

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...

80

Drilling fluids: Where should research dollars be spent  

Science Conference Proceedings (OSTI)

This article discusses the question of where to apply research dollars in the field of drilling fluids which is gravely impacted by environmental concerns. In fact, environmental regulations are the driving force in determining the thrust of drilling fluids research. For example, use of oil-base fluids offshore have, for all practical purposes, been precluded by high disposal costs since offshore disposal has been prohibited. Consequently it must be determined if a water-base mud can be developed that has all or most of the advantages of an oil-base mud.

Sauber

1987-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" 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

Research and Medical Isotope Reactor Supply | Y-12 National Security...  

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

Research and Medical ... Research and Medical Isotope Reactor Supply Our goal is to fuel research and test reactors with low-enriched uranium. Y-12 tops the short list of the...

82

Research Reactor Preparations for the Air Shipment of Highly Enriched Uranium from Romania  

SciTech Connect

In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation for conversion to low enriched uranium. The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR S research reactor at Magurele, Romania, to Chelyabinsk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation Rosatom and the International Atomic Energy Agency. Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel.

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

2010-03-01T23:59:59.000Z

83

Brookhaven Lab Completes Decommissioning of Graphite Research Reactor:  

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

Brookhaven Lab Completes Decommissioning of Graphite Research Brookhaven Lab Completes Decommissioning of Graphite Research Reactor: Reactor core and associated structures successfully removed; waste shipped offsite for disposal Brookhaven Lab Completes Decommissioning of Graphite Research Reactor: Reactor core and associated structures successfully removed; waste shipped offsite for disposal September 1, 2012 - 12:00pm Addthis The Brookhaven Graphite Research Reactor’s bioshield, which contains the 700-ton reactor core, is shown prior to decommissioning. The Brookhaven Graphite Research Reactor's bioshield, which contains the 700-ton reactor core, is shown prior to decommissioning. Pictured here is the Brookhaven Graphite Research Reactor, where major decommissioning milestones were recently reached after the remaining radioactive materials from the facility’s bioshield were shipped to a licensed offsite disposal facility.

84

Ames Laboratory Research Reactor Facility Ames, Iowa  

Office of Legacy Management (LM)

,, *' ; . Final Radiological Condition of the Ames Laboratory Research Reactor Facility Ames, Iowa _, . AGENCY: Office of Operational Safety, Department of Energy ' ACTION: Notice of Availability of Archival Information Package SUMMARY: The'Office of Operational Safety of the Department O i Energy (DOE) has reviewed documentation relating to the decontamination and decommissioning operations conducted at the Ames Laboratory Research Reactor Facility, Ames, Iowa and has prepared an archival informati0.n package to permanently document the results of the action and the site conditions and use restriction placed on the . site at the tim e of release. This review is based on post-decontamination survey data and other pertinent documentation referenced in and included in the archival package. The material and

85

Sterile Neutrino Search Using China Advanced Research Reactor  

E-Print Network (OSTI)

We study the feasibility of a sterile neutrino search at the China Advanced Research Reactor by measuring $\\bar {\

Guo, Gang; Ji, Xiangdong; Liu, Jianglai; Xi, Zhaoxu; Zhang, Huanqiao

2013-01-01T23:59:59.000Z

86

Reactor operations: Brookhaven Medical Research Reactor, Brookhaven High Flux Beam Reactor. Informal report, June 1995  

Science Conference Proceedings (OSTI)

Part one of this report gives the operating history of the Brookhaven Medical Research Reactor for the month of June. Also included are the BMRR technical safety surveillance requirements record and the summary of BMRR irradiations for the month. Part two gives the operating histories of the Brookhaven High Flux Beam Reactor and the Cold Neutron Facility at HFBR for June. Also included are the HFBR technical safety surveillance requirements record and the summary of HFBR irradiations for the month.

NONE

1995-06-01T23:59:59.000Z

87

Reactor operations: Brookhaven Medical Research Reactor, Brookhaven High Flux Beam Reactor. Informal report, July 1995  

Science Conference Proceedings (OSTI)

Part one of this report gives the operating history for the Brookhaven Medical Research Reactor for the month of July. Also included are the BMRR technical safety surveillance requirements record and the summary of BMRR irradiations for the month. Part two gives the operating histories for the Brookhaven High Flux Beam Reactor and the Cold Neutron Source Facility for the month of July. Also included are the HFBR technical safety surveillance requirements record and the summary of HFBR irradiations for the month.

NONE

1995-07-01T23:59:59.000Z

88

Chicago Pile reactors create enduring research legacy - Argonne's  

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

Chicago Pile reactors create enduring research Chicago Pile reactors create enduring research legacy 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

89

Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining  

SciTech Connect

A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) oxide fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs that were performed in succession with a single salt loading of molten LiCl – 1 wt% Li2O at 650 °C. Analysis of salt samples following the series of electrolytic reduction runs identified the diffusion of select fission products from the spent fuel to the molten salt electrolyte. The extents of metal oxide conversion in the post-test fuel were also quantified, including a nominal 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs that were performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500 °C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

S. D. Herrmann; S. X. Li

2010-09-01T23:59:59.000Z

90

Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program  

DOE Green Energy (OSTI)

Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions.

Not Available

1987-05-01T23:59:59.000Z

91

Research Reactors Division | ORNL Neutron Sciences  

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

Reactors Division (RRD) is responsible for operation of the High Flux Isotope Reactor (HFIR). Operating at 85 MW, HFIR is the highest flux reactor-based source of neutrons for...

92

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

93

A disposition strategy for highly enriched, aluminum-based fuel from research and test reactors  

SciTech Connect

The strategy proposed in this paper offers the Department of Energy an approach for disposing of aluminum-based, highly enriched uranium (HEU) spent fuels from foreign and domestic research reactors. The proposal is technically, socially, and economically sound. If implemented, it would advance US non-proliferation goals while also disposing of the spent fuel`s waste by timely and proven methods using existing technologies and facilities at SRS without prolonged and controversial storage of the spent fuel. The fuel would be processed through 221-H. The radioactive fission products (waste) would be treated along with existing SRS high level waste by vitrifying it as borosilicate glass in the Defense Waste Processing Facility (DWPF) for disposal in the national geological repository. The HEU would be isotopically diluted, during processing, to low-enriched uranium (LEU) which can not be used to make weapons, thus eliminating proliferation concerns. The LEU can be sold to fabricators of either research reactor fuel or commercial power fuel. This proposed processing-LEU recycle approach has several important advantages over other alternatives, including: Lowest capital investment; lowest net total cost; quickest route to acceptable waste form and final geologic disposal; and likely lowest safety, health, and environmental impacts.

McKibben, J.M.; Gould, T.H.; McDonell, W.R.; Bickford, W.E.

1994-11-01T23:59:59.000Z

94

Global estimation of potential unreported plutonium in thermal research reactors  

SciTech Connect

As of November, 1993, 303 research reactors (research, test, training, prototype, and electricity producing) were operational worldwide; 155 of these were in non-nuclear weapon states. Of these 155 research reactors, 80 are thermal reactors that have a power rating of 1 MW(th) or greater and could be utilized to produce plutonium. A previously published study on the unreported plutonium production of six research reactors indicates that a minimum reactor power of 40 MW (th) is required to make a significant quantity (SQ), 8 kg, of fissile plutonium per year by unreported irradiations. As part of the Global Nuclear Material Control Model effort, we determined an upper bound on the maximum possible quantity of plutonium that could be produced by the 80 thermal research reactors in the non-nuclear weapon states (NNWS). We estimate that in one year a maximum of roughly one quarter of a metric ton (250 kg) of plutonium could be produced in these 80 NNWS thermal research reactors based on their reported power output. We have calculated the quantity of plutonium and the number of years that would be required to produce an SQ of plutonium in the 80 thermal research reactors and aggregated by NNWS. A safeguards approach for multiple thermal research reactors that can produce less than 1 SQ per year should be conducted in association with further developing a safeguards and design information reverification approach for states that have multiple research reactors.

Dreicer, J.S.; Rutherford, D.A.

1996-09-01T23:59:59.000Z

95

Process and apparatus for recovery of fissionable materials from spent reactor fuel by anodic dissolution  

DOE Patents (OSTI)

An electrochemical process and apparatus for the recovery of uranium and plutonium from spent metal clad fuel pins is disclosed. The process uses secondary reactions between U.sup.+4 cations and elemental uranium at the anode to increase reaction rates and improve anodic efficiency compared to prior art processes. In another embodiment of the process, secondary reactions between Cd.sup.+2 cations and elemental uranium to form uranium cations and elemental cadmium also assists in oxidizing the uranium at the anode.

Tomczuk, Zygmunt (Orland Park, IL); Miller, William E. (Naperville, IL); Wolson, Raymond D. (Lockport, IL); Gay, Eddie C. (Park Forest, IL)

1991-01-01T23:59:59.000Z

96

Brookhaven Graphite Research Reactor Workshop | Department of Energy  

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

Services » Site & Facility Restoration » Deactivation & Services » Site & Facility Restoration » Deactivation & Decommissioning (D&D) » D&D Workshops » Brookhaven Graphite Research Reactor Workshop Brookhaven Graphite Research Reactor Workshop The Brookhaven Graphite Research Reactor (BGRR) was the first reactor built in the U.S. for peacetime atomic research following World War II. Construction began in 1947 and the reactor started operating in August 1950. In the next 18 years, an estimated 25,000 scientific experiments were carried out at the BGRR using neutrons produced in the facility's 700-ton graphite core, made up of more than 60,000 individual graphite blocks. The BGRR was placed on standby in 1968 and then permanently shut down as the next-generation reactor, the High Flux Beam Reactor (HFBR), was

97

The LEU conversion status of U.S. Research Reactors.  

SciTech Connect

This paper summarizes the conversion status of US research and test reactors and estimates uranium densities needed to convert reactors with power levels 21 MW from HEU ({ge} 20% U-235) to LEU (<20% U-235) fuels. Detailed conversion studies for each reactor need to be completed in order to establish the feasibility of using LEU fuels.

Matos, J. E.

1997-11-14T23:59:59.000Z

98

Brookhaven Graphite Research Reactor | Environmental Restoration Projects |  

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

Why Was the BGRR Decommissioned? Why Was the BGRR Decommissioned? BGRR The Brookhaven Graphite Research Reactor (BGRR) at Brookhaven National Laboratory (BNL) was decommissioned to ensure the complex is in a safe and stable condition and to reduce sources of groundwater contamination. The BGRR contained over 8,000 Curies of radioactive contaminants from past operations consisting of primarily nuclear activation products such as hydrogen-3 (tritium) and carbon-14 and fission products cesium-137 and strontium-90. The nature and extent of contamination varied by location depending on historic uses of the systems and components and releases, however, the majority of the contamination (over 99 percent) was bound within the graphite pile and biological shield. Radioactive contamination was identified in the fuel handling system deep

99

Brookhaven Graphite Research Reactor | Environmental Restoration Projects |  

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

- Cleanup Actions - Cleanup Actions Since the Brookhaven Graphite Research Reactor (BGRR) was shut down in 1968, many actions have been taken as part of the complex decommissioning. The actions undertaken throughout the BGRR complex ensure that the structures that remain are in a safe and stable condition and prepared it for long-term surveillance and maintenance. Regulatory Requirements The decommissioning of the BGRR was conducted under the federal Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). In 1992, an Interagency Agreement (PDF) among the DOE, the U.S. Environmental Protection Agency (EPA) and the New York State Department of Environmental Conservation (NYSDEC) became effective. The IAG provided the overall framework for conducting environmental restoration activities at

100

Computer-Based Model of the MIT Research Reactor  

SciTech Connect

A description is given of a model of the Massachusetts Institute of Technology Research Reactor (MITR) in which both the reactor's neutronic and thermal-hydraulic behaviors are replicated. The purpose of the model is to support control studies and the development of techniques for the automated diagnosis of reactivity transients. In particular, comparison of the model's predictions with actual measurements from the reactor will allow determination of whether the reactor is functioning as expected.

John A. Bernard; Lin-Wen Hu

2000-11-12T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" 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
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101

Opportunities for the Precision Study of Reactor Antineutrinos at Very Short Baselines at US Research Reactors  

E-Print Network (OSTI)

: pieter.mumm@nist.gov #12;2 past reactor experiments HFIR, ORNL NBSR, NIST ATR, INL available baselines at US research reactors 3 neutrino fit 3+1 neutrino fit Tuesday, August 7, 12 NIST ILL HFIR ATR SONGSNIST ILL HFIR ATR SONGS 10. 100 1000 core size reactor power reactorpower(MWth) 1meter ILL HFIR NBSR

102

Process and apparatus for recovery of fissionable materials from spent reactor fuel by anodic dissolution  

DOE Patents (OSTI)

An electrochemical process and apparatus for the recovery of uranium and plutonium from spent metal clad fuel pins is disclosed. The process uses secondary reactions between U{sup +4} cations and elemental uranium at the anode to increase reaction rates and improve anodic efficiency compared to prior art processes. In another embodiment of the process, secondary reactions between Cd{sup +2} cations and elemental uranium to form uranium cations and elemental cadmium also assists in oxidizing the uranium at the anode. 5 figs.

Tomczuk, Z.; Miller, W.E.; Wolson, R.D.; Gay, E.C.

1989-08-25T23:59:59.000Z

103

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

104

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research  

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

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program The Department of Energy's (DOE's) Light Water Reactor Sustainability (LWRS) Program is a five year effort that works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operation of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging

105

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors  

Science Conference Proceedings (OSTI)

The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

Rosenthal, Murray Wilford [ORNL

2009-08-01T23:59:59.000Z

106

Research on Very High Temperature Gas Reactors  

Science Conference Proceedings (OSTI)

Very high temperature gas reactors are helium-cooled, graphite-moderated advanced reactors that show potential for generating low-cost electricity via gas turbines or cogeneration with process-heat applications. This investigation addresses the development status of advanced coatings for nuclear-fuel particles and high-temperature structural materials and evaluates whether these developments are likely to lead to economically competitive applications of the very high temperature gas reactor concept.

1991-08-08T23:59:59.000Z

107

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

108

Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear  

National Nuclear Security Administration (NNSA)

Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear Maria Research Reactor loaded with LEU - Otwock, Poland | 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 > Media Room > Video Gallery > Maria Research Reactor loaded with LEU - ... Maria Research Reactor loaded with LEU - Otwock, Poland Maria Research Reactor loaded with LEU - Otwock, Poland

109

Brookhaven Graphite Research Reactor | Environmental Restoration...  

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

of multiple structures and systems that were necessary to operate and maintain the reactor. The most recognizable features of the complex include the Building 701 officeshigh...

110

RERTR program reduces use of enriched uranium in research reactors  

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

RERTR program reduces use of enriched uranium in research reactors RERTR program reduces use of enriched uranium in research reactors worldwide 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 RERTR program reduces use of enriched uranium in research reactors worldwide The High Flux Reactor in Petten, the Netherlands READY TO CONVERT - The High Flux Reactor in Petten, the Netherlands, has

111

Intermodal transfer of spent fuel  

Science Conference Proceedings (OSTI)

As a result of the international standardization of containerized cargo handling in ports around the world, maritime shipment handling is particularly uniform. Thus, handier exposure parameters will be relatively constant for ship-truck and ship-rail transfers at ports throughout the world. Inspectors' doses are expected to vary because of jurisdictional considerations. The results of this study should be applicable to truck-to-rail transfers. A study of the movement of spent fuel casks through ports, including the loading and unloading of containers from cargo vessels, afforded an opportunity to estimate the radiation doses to those individuals handling the spent fuels with doses to the public along subsequent transportation routes of the fuel. A number of states require redundant inspections and for escorts over long distances on highways; thus handlers, inspectors, escort personnel, and others who are not normally classified as radiation workers may sustain doses high enough to warrant concern about occupational safety. This paper addresses the question of radiation safety for these workers. Data were obtained during, observation of the offloading of reactor spent fuel (research reactor spent fuel, in this instance) which included estimates of exposure times and distances for handlers, inspectors and other workers during offloading and overnight storage. Exposure times and distance were also for other workers, including crane operators, scale operators, security personnel and truck drivers. RADTRAN calculational models and parameter values then facilitated estimation of the dose to workers during incident-free ship-to-truck transfer of spent fuel.

Neuhauser, K.S. (Sandia National Labs., Albuquerque, NM (United States)); Weiner, R.F. (Western Washington Univ., Bellingham, WA (United States))

1991-01-01T23:59:59.000Z

112

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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

113

Advanced Reactor Research and Development Funding Opportunity Announcement  

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

Reactor Research and Development Funding Opportunity Reactor Research and Development Funding Opportunity Announcement Advanced Reactor Research and Development Funding Opportunity Announcement The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) sponsors a program of research, development, and demonstration related to advanced non-light water reactor concepts. A goal of the program is to facilitate greater engagement between DOE and industry. During FY12, DOE established a Technical Review Panel (TRP) process to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D investment decisions. That process involved the use of a Request for Information (RFI) to solicit concept information from industry and engage technical experts to evaluate those concepts. Having completed this process, DOE desires to

114

Advanced Reactor Research and Development Funding Opportunity Announcement  

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

Advanced Reactor Research and Development Funding Opportunity Advanced Reactor Research and Development Funding Opportunity Announcement Advanced Reactor Research and Development Funding Opportunity Announcement The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) sponsors a program of research, development, and demonstration related to advanced non-light water reactor concepts. A goal of the program is to facilitate greater engagement between DOE and industry. During FY12, DOE established a Technical Review Panel (TRP) process to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D investment decisions. That process involved the use of a Request for Information (RFI) to solicit concept information from industry and engage technical experts to evaluate those concepts. Having completed this process, DOE desires to

115

Present Status Of Research Reactor Decommissioning Program In Indonesia  

E-Print Network (OSTI)

At present, Indonesia has 3 research reactors: MTR-type multipurpose reactor of 30 MW at Serpong site, TRIGA-type research reactor of 1 MW at Bandung site, and small TRIGA - type reactor of 100 kW at Yogyakarta Research Center. The oldest one is the TRIGA reactor at Bandung site, which went critical at 250 kW in 1964, then was operated at maximum of 1000 kW by 1971. The reactor has operated for a total of 35 years. There is no decision for decommissioning this reactor; however, slowly but surely, it will be an object for a near-future decommissioning program. Anticipation of the situation is necessary. For the Indonesian case, early decommissioning strategy for a research reactor and restricted use of the site for another nuclear installation is favorable under high land pricing, availability of radwaste repository, and cost analysis. Graphite from Triga reactor reflector is recommended for direct disposal after conditioning, without volume reduction treatment. Development of human ...

Mulyanto And Gunandjar

2000-01-01T23:59:59.000Z

116

Sandia National Laboratories: Research: Facilities: Sandia Pulsed Reactor  

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

Sandia Pulsed Reactor Facility - Critical Experiments Sandia Pulsed Reactor Facility - Critical Experiments Sandia scientist John Ford places fuel rods in the Seven Percent Critical Experiment (7uPCX) at the Sandia Pulsed Reactor Facility Critical Experiments (SPRF/CX) test reactor - a reactor stripped down to its simplest form. The Sandia Pulsed Reactor Facility - Critical Experiments (SPRF/CX) provides a flexible, shielded location for performing critical experiments that employ different reactor core configurations and fuel types. The facility is also available for hands-on nuclear criticality safety training. Research and other activities The 7% series, an evaluation of various core characteristics for higher commercial-fuel enrichment, is currently under way at the SPRF/CX. Past critical experiments at the SPRF/CX have included the Burnup Credit

117

Studies and research concerning BNFP design and construction of a spent-fuel disassembly/encapsulation system  

SciTech Connect

Commercial light water power reactor operation in the United States is developing a cumbersome inventory of spent fuel. Systems for interim storage and handling of this fuel are being developed by the Federal Government and industry. Disassembly and canning of the spent fuel elements is one of these systems. It has the potential to double the storage capacity of a prereprocessing storage facility or to triple the capacity of conventional shipping casks. Prototype equipment and controls required to perform this operation in a dry environment have been primarily designed and fabricated at the Barnwell Nuclear Fuel Plant (BNFP). Ridihalgh, Eggers, and Associates have provided design support and fabrication of the control system. This system is capable of extracting and canning the fuel pins and compacting the nonfuel-bearing components of spent fuel assemblies at processing rates of 10 to 12 assemblies per day. The process also provides the potential for enhanced inspection and assay of spent fuel by reducing the interference encountered from the high gamma fields of the nonfuel bearing hardware.

Dabolt, R.J.

1981-04-01T23:59:59.000Z

118

Sodium fast reactor safety and licensing research plan. Volume II.  

SciTech Connect

Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d'%C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

2012-05-01T23:59:59.000Z

119

Engineering activities at the MIT research reactor in support of power reactor technology  

SciTech Connect

The Massachusetts Institute of Technology (MIT) research reactor (MITR-II) is a 5-MW(thermal) light-water-cooled and-moderated reactor (LWR) with in-core neutron and gamma dose rates that closely approximate those in current LWRs. Compact in-pile loops that simulate pressurized water reactor (PWR) and boiling water reactor (BWR) thermal hydraulics and coolant chemistry have been designed for installation in the MITR-II. A PWR loop has been completed and is currently operating in the reactor. A BWR loop is under construction, and an in-pile facility for irradiation-assisted stress corrosion crack (IASCC) testing is being designed. Another major area of research and on-line testing is the closed-loop, nonlinear, digital control of various reactor parameters, including the power level, temperature, and net energy production.

Harling, O.K.; Bernard, J.A.; Driscoll, M.J.; Kohse, G.E.; Ballinger, R.G.

1989-01-01T23:59:59.000Z

120

Evaluation of cover gas impurities and their effects on the dry storage of LWR (light-water reactor) spent fuel  

DOE Green Energy (OSTI)

The purposes of this report are to (1) identify the sources of impurity gases in spent fuel storage casks; (2) identify the expected concentrations and types of reactive impurity gases from these sources over an operating lifetime of 40 years; and (3) determine whether these impurities could significantly degrade cladding or exposed fuel during this period. Four potential sources of impurity gases in the helium cover gas in operating casks were identified and evaluated. Several different bounding cases have been considered, where the reactive gas inventory is either assumed to be completely gettered by the cladding or where all oxygen is assumed to react completely with the exposed fuel. It is concluded that the reactive gas inventory will have no significant effect on the cladding unless all available oxygen reacts with the UO/sub 2/ fuel to produce U/sub 3/O/sub 8/ at one or two cladding breaches. Based on Zircaloy oxidation data, the oxygen inventory in a fully loaded pressurized water reactor cask such as the Castor-V/21 will be gettered by the Zircaloy cladding in about 1 year if the peak cladding temperature within the task is greater than or equal to300/sup 0/C. Only a negligible decrease in the thickness of the cladding would result. 24 refs., 4 tabs.

Knoll, R.W.; Gilbert, E.R.

1987-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" 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

Automated startup of the MIT research reactor  

SciTech Connect

This summary describes the development, implementation, and testing of a generic method for performing automated startups of nuclear reactors described by space-independent kinetics under conditions of closed-loop digital control. The technique entails first obtaining a reliable estimate of the reactor's initial degree of subcriticality and then substituting that estimate into a model-based control law so as to permit a power increase from subcritical on a demanded trajectory. The estimation of subcriticality is accomplished by application of the perturbed reactivity method. The shutdown reactor is perturbed by the insertion of reactivity at a known rate. Observation of the resulting period permits determination of the initial degree of subcriticality. A major advantage to this method is that repeated estimates are obtained of the same quantity. Hence, statistical methods can be applied to improve the quality of the calculation.

Kwok, K.S. (Massachusetts Inst. of Tech., Cambridge (United States))

1992-01-01T23:59:59.000Z

122

Argonne's pyroprocessing and advanced reactor research featured on WGN  

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

Argonne's pyroprocessing and advanced reactor research featured on WGN Argonne's pyroprocessing and advanced reactor research featured on WGN radio 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 Argonne's pyroprocessing and advanced reactor research featured on WGN radio Uranium dendrites These tiny branches, or "dendrites," of pure uranium form when engineers

123

Nuclear reactor and materials science research: Final technical report, May 1, 1985-September 30, 1986. [Academic and research utilization of reactor  

SciTech Connect

Throughout the 17-month period of the grant, May 1, 1985 - September 30, 1986, the MIT Research Reactor (MITR-II) was operated in support of research and academic programs in the physical and life sciences and in related engineering fields. The period encompassed MIT's fiscal year utilization of the reactor during that period may be classified as follows: neutron beam tube research, nuclear materials research and development, radiochemistry and trace analysis, nuclear medicine, radiation health physics, computer control of reactors, dose reduction in nuclear power reactors, reactor irradiations and services for groups outside MIT, and MIT research reactor. This paper provides detailed information on this research academic utilization.

Harling, O.K.

1987-05-11T23:59:59.000Z

124

Spent Fuel Background Report Volume I  

Science Conference Proceedings (OSTI)

This report is an overview of current spent nuclear fuel management in the DOE complex. Sources of information include published literature, internal DOE documents, interviews with site personnel, and information provided by individual sites. Much of the specific information on facilities and fuels was provided by the DOE sites in response to the questionnaire for data for spent fuels and facilities data bases. This information is as accurate as is currently available, but is subject to revision pending results of further data calls. Spent fuel is broadly classified into three categories: (a) production fuels, (b) special fuels, and (c) naval fuels. Production fuels, comprising about 80% of the total inventory, are those used at Hanford and Savannah River to produce nuclear materials for defense. Special fuels are those used in a wide variety of research, development, and testing activities. Special fuels include fuel from DOE and commercial reactors used in research activities at DOE sites. Naval fuels are those developed and used for nuclear-powered naval vessels and for related research and development. Given the recent DOE decision to curtail reprocessing, the topic of main concern in the management of spent fuel is its storage. Of the DOE sites that have spent nuclear fuel, the vast majority is located at three sites-Hanford, INEL, and Savannah River. Other sites with spent fuel include Oak Ridge, West Valley, Brookhaven, Argonne, Los Alamos, and Sandia. B&W NESI Lynchburg Technology Center and General Atomics are commercial facilities with DOE fuel. DOE may also receive fuel from foreign research reactors, university reactors, and other commercial and government research reactors. Most DOE spent fuel is stored in water-filled pools at the reactor facilities. Currently an engineering study is being performed to determine the feasibility of using dry storage for DOE-owned spent fuel currently stored at various facilities. Delays in opening the deep geologic repository and the decision to phase out reprocessing of production fuels are extending the need for interim storage. The report describes the basic storage conditions and the general SNF inventory at individual DOE facilities.

Abbott, D.

1994-03-01T23:59:59.000Z

125

DISMANTLING OF THE REACTOR BLOCK OF THE FRJ-1 RESEARCH REACTOR (MERLIN)  

SciTech Connect

This report describes the past procedure in dismantling the reactor block of the FRJ-1 research reactor (MERLIN). Furthermore, it gives an outlook on future activities up to the final removal of the reactor block. MERLIN is an abbreviation for Medium Energy Research Light Water Moderated Industrial Nuclear Reactor. The FRJ-1 (MERLIN) was shut down in 1985 and the fuel elements removed from the facility. After dismantling the coolant loops and removing the reactor tank internals with subsequent draining of the reactor tank water, the first activities for dismantling the reactor block were carried out in summer 2001. The relevant license was granted in late July 2001 by the licensing authority specifying 8 incidental provisions. After dismantling the reactor extension (gates of the thermal columns and steel platforms surrounding the reactor block), a heavy-load platform including a casing around the reactor block was constructed. Two ventilation systems with a volume flow of 10,000 and 2 ,000 m3/h will, moreover, serve to avoid a spread of contamination. The reactor block will be dismantled in three phases divided according to upper, central and bottom sections. Dismantling the upper section started in August 2002. This section as well as the bottom section can probably be completely measured for clearance. For this reason, the activities have so far been carried out manually using mechanical and thermal techniques. The central section will probably have to be largely disposed of as radioactive waste. This is the region of the former reactor core in which the experimental devices are also integrated. Most of this work will probably have to be carried out by remote handling. More than 80 % of the dismantled materials of the reactor block can probably be measured for clearance. For this purpose, a clearance measurement device was taken into operation in the FRJ-1. On this occasion, the limits of clearance measurement have become evident. For concrete, which constitutes the largest portion of the dismantled materials by volume, an additional conditioning step has become necessary to fulfill the clearance criteria, whereas waste packages with steel components largely have to be reconditioned once more at a later stage. Material measured for clearance will be disposed of conventionally (recycling, landfill) after inspection by the official expert and clearance by the regulatory authority. Dismantled parts that cannot be measured for clearance will be transferred to the Decontamination Department of the Research Centre. From the present perspective, the dismantling of the reactor block will be completed within the first six months of 2003.

Stahn, B.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

2003-02-27T23:59:59.000Z

126

LEU conversion status of US research reactors, September 1996  

SciTech Connect

This paper summarizes the conversion status of research and test reactors in the United States from the use of fuels containing highly- enriched uranium (HEU, greater than or equal to 20%) to the use of fuels containing low-enriched uranium (LEU, < 20%). Estimates of the uranium densities required for conversion are made for reactors with power levels greater than or equal to 1 MW that are not currently involved in the LEU conversion process.

Matos, J.E.

1996-10-07T23:59:59.000Z

127

Reduced enrichment for research and test reactors: Proceedings  

SciTech Connect

The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

1988-05-01T23:59:59.000Z

128

Research reactor usage at the Idaho National Engineering Laboratory in support of university research and education  

SciTech Connect

The Idaho National Engineering Laboratory is a US Department of Energy laboratory which has a substantial history of research and development in nuclear reactor technologies. There are a number of available nuclear reactor facilities which have been incorporated into the research and training needs of university nuclear engineering programs. This paper addresses the utilization of the Advanced Reactivity Measurement Facility (ARMF) and the Coupled Fast Reactivity Measurement Facility (CFRMF) for thesis and dissertation research in the PhD program in Nuclear Science and Engineering by the University of Idaho and Idaho State University. Other reactors at the INEL are also being used by various members of the academic community for thesis and dissertation research, as well as for research to advance the state of knowledge in innovative nuclear technologies, with the EBR-II facility playing an essential role in liquid metal breeder reactor research. 3 refs.

Woodall, D.M.; Dolan, T.J.; Stephens, A.G. (Idaho National Engineering Lab., Idaho Falls, ID (USA))

1990-01-01T23:59:59.000Z

129

Determining Plutonium Mass in Spent Fuel with Nondestructive Assay Techniques NGSI Research Overview and Update on NDA Techniques  

E-Print Network (OSTI)

Determining Plutonium Mass in Spent Fuel with Non-CN-184/137 Determining Plutonium Mass in Spent Fuel withthe Direct Measurement of Plutonium in Spent LWR Fuels by

A., V. Mozin, S.J. Tobin, L.W. Cambell, J.R. Cheatham, C.R. Freeman, C.J. Gesh,

2012-01-01T23:59:59.000Z

130

Reactor Safety Research: Semiannual report, July-December 1986  

DOE Green Energy (OSTI)

Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the tehcnology base supporting licensing decisions.

Not Available

1987-11-01T23:59:59.000Z

131

Measurement of Spent Fuel Assemblies - Overview of the Status of the Technology for Initiating Discussion at NATIONAL RESEARCH CENTRE KURCHATOV INSTITUTE June 2013  

SciTech Connect

This presentation provides an overview of the status of the technology for the measurement of the fissile material content of spent nuclear reactor fuel. The emphasis is on how the needs of the U.S. Nuclear Regulatory Commission and the International Atomic Energy Agency are met by the available technology and what more needs to be done in this area.

SISKIND B.; N /A

2013-06-03T23:59:59.000Z

132

New Research Center to Increase Safety and Power Output of U.S. Nuclear Reactors  

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

The Department of Energy dedicated the Consortium for Advanced Simulation of Light Water Reactors (CASL), an advanced research facility that will accelerate the advancement of nuclear reactor technology.

133

RERTR 2009 (Reduced Enrichment for Research and Test Reactors)  

SciTech Connect

The U.S. Department of Energy/National Nuclear Security Administration's Office of Global Threat Reduction in cooperation with the China Atomic Energy Authority and International Atomic Energy Agency hosted the 'RERTR 2009 International Meeting on Reduced Enrichment for Research and Test Reactors.' The meeting was organized by Argonne National Laboratory, China Institute of Atomic Energy and Idaho National Laboratory and was held in Beijing, China from November 1-5, 2009. This was the 31st annual meeting in a series on the same general subject regarding the conversion of reactors within the Global Threat Reduction Initiative (GTRI). The Reduced Enrichment for Research and Test Reactors (RERTR) Program develops technology necessary to enable the conversion of civilian facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets.

Totev, T.; Stevens, J.; Kim, Y. S.; Hofman, G.; Matos, J.; Hanan, N.; Garner, P.; Dionne, B.; Olson, A.; Feldman, E.; Dunn, F.; Nuclear Engineering Division; Atomic Research Center; Inst. of Nuclear Physics; LLNL; INL; Korea Atomic Energy Research Inst.; Comisi?n Nacional de Energ?a At?mica; Nuclear Reactor Lab.; Inst. of Atomic Energy-Poland; AECL-Canada; Hungarian Academy of Sciences KFKI Atomic Energy Research Inst.; Japan Atomic Energy Agency; Nuclear Power Inst. of China; Kyoto Univ. Research Reactor Inst.

2010-03-01T23:59:59.000Z

134

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

135

Reactor Safety Research Programs Quarterly Report October - December 1980  

SciTech Connect

This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from October 1 through December 31, 1980, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NOE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Edler, S K

1981-04-01T23:59:59.000Z

136

Processing of FRG high-temperature gas-cooled reactor fuel elements at General Atomic under the US/FRG cooperative agreement for spent fuel elements  

Science Conference Proceedings (OSTI)

The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects of gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment development section of the agreement, both FRG mixed uranium/ thorium and low-enriched uranium fuel spheres have been processed in the Department of Energy-sponsored cold pilot plant for high-temperature gas-cooled reactor (HTGR) fuel processing at General Atomic Company in San Diego, California. The FRG fuel spheres were crushed and burned to recover coated fuel particles suitable for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated certain modifications to the US HTGR fuel burining process necessary for FRG fuel treatment. Results of the tests will be used in the design of a US/FRG joint prototype headend facility for HTGR fuel.

Holder, N.D.; Strand, J.B.; Schwarz, F.A.; Drake, R.N.

1981-11-01T23:59:59.000Z

137

Evaluation of potential food applications of dried distillers spent grain (DSG). Final research report  

SciTech Connect

Results from experimental test bakes indicate that dried distillers spent grain (DSG) can be used to replace up to 15% of the flour for the production of an acceptable variety bread, provided that the DSG is processed under optimum conditions for a satisfactory flavor and color development. The raw materials used for the mash bill may also affect the taste of the finished product. (MHR)

Not Available

1981-02-06T23:59:59.000Z

138

Control of degradation of spent LWR (light-water reactor) fuel during dry storage in an inert atmosphere  

DOE Green Energy (OSTI)

Dry storage of Zircaloy-clad spent fuel in inert gas (referred to as inerted dry storage or IDS) is being developed as an alternative to water pool storage of spent fuel. The objectives of the activities described in this report are to identify potential Zircaloy degradation mechanisms and evaluate their applicability to cladding breach during IDS, develop models of the dominant Zircaloy degradation mechanisms, and recommend cladding temperature limits during IDS to control Zircaloy degradation. The principal potential Zircaloy cladding breach mechanisms during IDS have been identified as creep rupture, stress corrosion cracking (SCC), and delayed hydride cracking (DHC). Creep rupture is concluded to be the primary cladding breach mechanism during IDS. Deformation and fracture maps based on creep rupture were developed for Zircaloy. These maps were then used as the basis for developing spent fuel cladding temperature limits that would prevent cladding breach during a 40-year IDS period. The probability of cladding breach for spent fuel stored at the temperature limit is less than 0.5% per spent fuel rod. 52 refs., 7 figs., 1 tab.

Cunningham, M.E.; Simonen, E.P.; Allemann, R.T.; Levy, I.S.; Hazelton, R.F.

1987-10-01T23:59:59.000Z

139

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

140

Reactor core design and modeling of the MIT research reactor for conversion to LEU  

SciTech Connect

Feasibility design studies for conversion of the MIT Research Reactor (MITR) to LEU are described. Because the reactor fuel has a rhombic cross section, a special input processor was created in order to model the reactor in great detail with the REBUS-PC diffusion theory code, in 3D (triangular-z) geometry. Comparisons are made of fuel assembly power distributions and control blade worth vs. axial position, between REBUS-PC results and Monte Carlo predictions from the MCNP code. Results for the original HEU core at zero burnup are also compared with measurement. These two analysis methods showed remarkable agreement. Ongoing fuel cycle studies are summarized. A status report will be given as to results thus far that affect key design decisions. Future work plans and schedules to achieve completion of the conversion are presented. (author)

Newton, Thomas H. Jr. [Nuclear Reactor Laboratory, Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Olson, Arne P.; Stillman, John A. [RERTR Program, Argonne National Laboratory, Argonne, IL 60439 (United States)

2008-07-15T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" from the National Library of EnergyBeta (NLEBeta).
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141

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

142

Two U.S. University Research Reactors to be Converted From Highly...  

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

are here Home Two U.S. University Research Reactors to be Converted From Highly Enriched Uranium to Low-Enriched Uranium Two U.S. University Research Reactors to be Converted...

143

Design and optimization of a high thermal flux research reactor via Kriging-based algorithm  

E-Print Network (OSTI)

In response to increasing demands for the services of research reactors, a 5 MW LEU-fueled research reactor core is developed and optimized to provide high thermal flux within specified limits upon thermal hydraulic ...

Kempf, Stephanie Anne

2011-01-01T23:59:59.000Z

144

Friction pressure drop measurements and flow distribution analysis for LEU conversion study of MIT Research Reactor  

E-Print Network (OSTI)

The MIT Nuclear Research Reactor (MITR) is the only research reactor in the United States that utilizes plate-type fuel elements with longitudinal fins to augment heat transfer. Recent studies on the conversion to low-enriched ...

Wong, Susanna Yuen-Ting

2008-01-01T23:59:59.000Z

145

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

146

Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies Phase I Study  

SciTech Connect

A novel methodology to detect diversion of spent fuel from Pressurized Water Reactors (PWR) has been developed in order to address a long unsolved safeguards verification problem for international safeguards community such as International Atomic Energy Agency (IAEA) or European Atomic Energy Community (EURATOM). The concept involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed fuel assembly. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the assembly will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard laptop computers. Monte Carlo simulation studies and a set of controlled experiments with actual commercial PWR spent fuel assemblies were performed and validated this novel methodology. Based on the simulation studies and benchmarking measurements, the methodology developed promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the detection threshold of 50% missing pins from a spent fuel assembly, a threshold defined by the IAEA Safeguards Criteria. The methodology does not rely on any operator provided data like burnup or cooling time and does not require movement of the fuel assembly from the storage rack in the spent fuel pool. A concept was developed to build a practical field device, Partial Defect Detector (PDET), which will be completely portable and will use standard radiation measuring devices already in use at the IAEA. The use of the device will not require any information provided by the operator or any prior knowledge of the spent fuel assembly. The device can also be operated without any movement of the spent fuel from its storage position. Based on parametric studies conducted via computer simulation, the device should be able to detect diversion of as low as ten percent of the missing or replaced fuel from an assembly regardless of the location of the missing fuel within the assembly, of the cooling time, initial fuel enrichment or burnup levels. Conditions in the spent fuel pool such as clarity of the water or boron content are also not issues for this device. The shape of the base signature is principally dependent on the layout of the guide tubes in the various types of PWR fuel assemblies and perturbations in the form of replaced fuel pins will distort this signature. These features of PDET are all unique and overcome limitation and disadvantages presented by currently used devices such as the Fork detector or the Cerenkov Viewing Device. Thus, this device when developed and tested could fill an important need in the safeguards area for partial defect detection, a technology that the IAEA has been seeking for the past few decades.

Ham, Y S; Sitaraman, S

2008-12-24T23:59:59.000Z

147

Decommissioning of the Austrian 10 MW Research Reactor, Results and Lessons learned Paper  

SciTech Connect

After the decision to shut down the 10 MW ASTRA-MTR Research Reactor was reached in May 1998, the possible options and required phases for decommissioning and removal of the radioactive components were evaluated in a decommissioning study. To support the decisions at each phase, an estimate of the activity inventory in the various parts of the reactor and the waste volume to be expected was performed. Of the possible options an immediate dismantling to phase 1 of IAEA Technical Guide Lines after the immediately following, continued dismantling to phase 2 of these guide lines was identified as the most reasonable and under the auspices optimum choice. The actual decommissioning work on the ASTRA-Reactor began in January 2000 after its final shutdown on July 31, 1999. Preliminary evaluations of the activity inventory gave an estimated amount of 320 kg of intermediate level waste, of about 60 metric tons of contaminated and another 100 metric tons of activated low level radioactive waste. The activities were roughly estimated to be at 200 TBq in the intermediate level and 6 GBq in the low level. The structure of the decommissioning process was decided against cost-, time- and risk-optimization following the basic layout of the main tasks, e.g. the removing of the fuel, the recovering and the treatment of the intermediate level activities in the vicinity of the core, the handling and conditioning of the neutron exposed graphite and the Beryllium-elements. As an example, the dismantling of approx. 1400 metric tons of the biological shield is described in more detail from the determination of the dismantling technique to the clearing procedures and the deposition. The process of dismantling of the biological shield is presented in fast motion. The dismantling of the pump-room installations of the primary loop, the processing of the contaminated or activated metals, the dismantling of the ventilation system and the radiological clearance of the reactor building was done under optimized conditions and is explained in the following. Spent fuel was generally delivered to the US Department of Energy - DOE in several shipments over the operational time of the ASTRA reactor. With the last shipment in May 2001 all the remaining spent fuel elements out of the ASTRA reactor consignment were transferred to DOE. To reduce waste from concrete shielding, German regulations Dt.StrSchV, annex IV, table 1, two clearance values referring to 'clearance restricted for permanent deposit' and to a clearance for unrestricted re-use were used. In order to reduce the amount of an estimated 60 tons of slightly contaminated metals, it was determined that introducing re-melting procedures were the most economical way. To obtain radiological clearance of the reactor building, compliance with the release limits according to Austrian Radiation Protection Ordinance had to be proved to the regulatory body. There, in general, the limits for unrestricted release were defined as a maximum dose rate of 10 {mu}Sv effective for an individual person per year. The results of the regular yearly medical examinations of the staff indicated no influence of the work related to decommissioning. The readings of the personal dosimeters over the entire project amounted to a total of 85.6 mSv, averaging to 1.07 mSv per year and person. After finishing the decommissioning process, the material balance showed 89.6 % for unrestricted reuse, 6.6 % for conventional mass-dumping and 3.8 % of ILW and LLW. The project was covered by an extensive documentation. All operations within NES followed ISO 9000 quality insurance standards. Experiences and knowledge were presented to and shared with the community, e.g. AFR and IAEA throughout the project. (authors)

Hillebrand, G.; Meyer, F. [Nuclear Engineering Seibersdorf GmbH (NES), Seibersdorf, Austria, Europe (Austria)

2008-07-01T23:59:59.000Z

148

Regulatory Experiences for the Decommissioning of the Research Reactor in Korea  

Science Conference Proceedings (OSTI)

The first research reactor in Korea (KRR-1, TRIGA Mark-II) has operated since 1962, and the second one (KRR-2, TRIGA Mark-III), since 1972. Both of them were phased out in 1995 due to their lives and the operation of a new research reactor, HANARO (30 MW thermal power) operated by KAERI (Korea Atomic Energy Research Institute). After deciding the shutdown by the Nuclear Development and Utilization Committee in March 1996, KAERI began to prepare the decommissioning plan, including the environmental impact assessment, and submitted the plan to the Ministry of Science and Technology (MOST) in December 1998. Korea Institute of Nuclear Safety (KINS) reviewed document and prepared the review report in 1999. KINS is an organization of technical expertise which performs regulatory functions, entrusted by the MOST in accordance with the Atomic Energy Act and its Enforcement Decree. The review report written by KINS was consulted by the Special Committee on Nuclear Safety in January 2000. The committee submitted their consultation results to the Nuclear Safety Commission for the final approval by the Minister of MOST. The license was issued in November 2000. With the consent of the Korean government to the US Record of Decision, the spent fuel of KRR-1 and 2 was safely transported to the United States in July 1998. The decontamination and dismantling of KRR-2 was completed at the end of 2005 but the decommissioning of KRR-1 has been suspended by the problem for the memorial of the reactor. After the decommissioning of the research reactor is finished, the site will be returned to the site owner, Korea Electric Power Corporation (KEPCO). In this paper, the state-of-art and lessons learnt from recent regulatory activities for decommissioning of KRR- 2 are summarized. In conclusion: since the shutdown of KRR-1 and 2 had been decided, the safe assessment and licensing review were carried out after applying for decommissioning plan of those research reactors by operator. Through the safety assessment and license review, the plan was approved in 2000. The D and D of KRR-2 except Reactor hall of KRR-1 has been performed safely and completed in 2005. On the other hand the integrated safety for the decommissioning has been confirmed by regulatory team inspection. The radioactive waste arising from the dismantling work has been packed mainly in a 4 m{sup 3} containers and stored on site, reactor hall of KRR-2, until the LILW disposal facility is operational. For the whole dismantling work, with respect to the ALARA principle, not only worker protection from radiation and industrial hazard, but also the environment protection should be the first priority. Also much attention has been paid to the record keeping for the future decommissioning of nuclear power plants.

CHOI, Kyung-Woo [Korea Institute of Nuclear Safety, P.O Box 114, Yuseong, Daejeon (Korea, Republic of)

2008-01-15T23:59:59.000Z

149

Design Calculations of VPVR/M Research Reactor Spent-Fuel Cask  

Science Conference Proceedings (OSTI)

Miscellaneous / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (PART 3) / Radioactive Waste Management and Disposal

Milan Marek

150

Advanced sodium fast reactor accident source terms : research needs.  

Science Conference Proceedings (OSTI)

An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

Powers, Dana Auburn; Clement, Bernard [IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France] IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France; Denning, Richard [Ohio State University, Columbus, OH] Ohio State University, Columbus, OH; Ohno, Shuji [Japan Atomic Energy Agency, Ibaraki, Japan] Japan Atomic Energy Agency, Ibaraki, Japan; Zeyen, Roland [Institute for Energy Petten, Saint-Paul-lez-Durance, France] Institute for Energy Petten, Saint-Paul-lez-Durance, France

2010-09-01T23:59:59.000Z

151

Advanced sodium fast reactor accident source terms : research needs.  

SciTech Connect

An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

Powers, Dana Auburn; Clement, Bernard [IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France] IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France; Denning, Richard [Ohio State University, Columbus, OH] Ohio State University, Columbus, OH; Ohno, Shuji [Japan Atomic Energy Agency, Ibaraki, Japan] Japan Atomic Energy Agency, Ibaraki, Japan; Zeyen, Roland [Institute for Energy Petten, Saint-Paul-lez-Durance, France] Institute for Energy Petten, Saint-Paul-lez-Durance, France

2010-09-01T23:59:59.000Z

152

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

153

Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core  

Science Conference Proceedings (OSTI)

The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried out at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.

Bokhari, Ishtiaq H. [Pakistan Institute of Nuclear Science and Technology (Pakistan)

2004-12-15T23:59:59.000Z

154

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

155

Eastern Europe Research Reactor Initiative nuclear education and training courses - Current activities and future challenges  

Science Conference Proceedings (OSTI)

The Eastern Europe Research Reactor Initiative was established in January 2008 to enhance cooperation between the Research Reactors in Eastern Europe. It covers three areas of research reactor utilisation: irradiation of materials and fuel, radioisotope production, neutron beam experiments, education and training. In the field of education and training an EERRI training course was developed. The training programme has been elaborated with the purpose to assist IAEA Member States, which consider building a research reactor (RR) as a first step to develop nuclear competence and infrastructure in the Country. The major strength of the reactor is utilisation of three different research reactors and a lot of practical exercises. Due to high level of adaptability, the course can be tailored to specific needs of institutions with limited or no access to research reactors. (authors)

Snoj, L. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Sklenka, L.; Rataj, J. [Dept. of Nuclear Reactor, Czech Technical Univ. in Prague, V Holesovickach 2, 180 00 Prague 8 (Czech Republic); Boeck, H. [Vienna Univ. of Technology/Atominstitut, Stadionallee 2, 1020 Vienna (Austria)

2012-07-01T23:59:59.000Z

156

Recycling contaminated spent blasting abrasives in Portland cement mortars using solidification/stabilization technology. Final research report  

SciTech Connect

The use of abrasive blasting to remove lead-based paint from steel bridges produces contaminated spent blasting abrasives that may be classified as hazardous by the Environmental Protection Agency. If the spent abrasives leach lead, chromium, and cadmium greater than the EPA limits, the spent abrasives are classified as hazardous and must be rendered nonhazardous prior to disposal. An alternative to disposing of spent blasting abrasives is to recycle them in an environmentally sound application. This study investigates the feasibility of recycling spent blasting abrasives in portland cement mortars using solidification/stabilization technology to produce a usable construction material. Field application guidelines for recycling spent blasting abrasives in portland cement mortars are established based upon the results of this study.

Salt, B.K.; Carrasquillo, R.L.; Loehr, R.C.; Fowler, D.W.

1995-04-01T23:59:59.000Z

157

Role of research reactors in training of NPP personnel with special focus on training reactor VR-1  

Science Conference Proceedings (OSTI)

Research reactors play an important role in providing key personnel of nuclear power plants a hands-on experience from operation and experiments at nuclear facilities. Training of NPP (Nuclear Power Plant) staff is usually deeply theoretical with an extensive utilisation of simulators and computer visualisation. But a direct sensing of the reactor response to various actions can only improve the personnel awareness of important aspects of reactor operation. Training Reactor VR-1 and its utilization for training of NPP operators and other professionals from Czech Republic and Slovakia is described. Typical experimental exercises and good practices in organization of a training program are demonstrated. (authors)

Sklenka, L.; Rataj, J.; Frybort, J.; Huml, O. [Dept. of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical Univ. in Prague, V Holesovickach 2, Prague 8, 180 00 (Czech Republic)

2012-07-01T23:59:59.000Z

158

Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 1, Activation measurements and comparison with calculations for spent fuel assembly hardware  

Science Conference Proceedings (OSTI)

Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1. 5 refs., 4 figs., 21 tabs.

Luksic, A.

1989-06-01T23:59:59.000Z

159

Modular Pebble Bed Reactor Project, University Research Consortium Annual Report  

Science Conference Proceedings (OSTI)

This project is developing a fundamental conceptual design for a gas-cooled, modular, pebble bed reactor. Key technology areas associated with this design are being investigated which intend to address issues concerning fuel performance, safety, core neutronics and proliferation resistance, economics and waste disposal. Research has been initiated in the following areas: · Improved fuel particle performance · Reactor physics · Economics · Proliferation resistance · Power conversion system modeling · Safety analysis · Regulatory and licensing strategy Recent accomplishments include: · Developed four conceptual models for fuel particle failures that are currently being evaluated by a series of ABAQUS analyses. Analytical fits to the results are being performed over a range of important parameters using statistical/factorial tools. The fits will be used in a Monte Carlo fuel performance code, which is under development. · A fracture mechanics approach has been used to develop a failure probability model for the fuel particle, which has resulted in significant improvement over earlier models. · Investigation of fuel particle physio-chemical behavior has been initiated which includes the development of a fission gas release model, particle temperature distributions, internal particle pressure, migration of fission products, and chemical attack of fuel particle layers. · A balance of plant, steady-state thermal hydraulics model has been developed to represent all major components of a MPBR. Component models are being refined to accurately reflect transient performance. · A comparison between air and helium for use in the energy-conversion cycle of the MPBR has been completed and formed the basis of a master’s degree thesis. · Safety issues associated with air ingress are being evaluated. · Post shutdown, reactor heat removal characteristics are being evaluated by the Heating-7 code. · PEBBED, a fast deterministic neutronic code package suitable for numerous repetitive calculations has been developed. Use of the code has focused on scoping studies for MPBR design features and proliferation issues. Publication of an archival journal article covering this work is being prepared. · Detailed gas reactor physics calculations have also been performed with the MCNP and VSOP codes. Furthermore, studies on the proliferation resistance of the MPBR fuel cycle has been initiated using these code · Issues identified during the MPBR research has resulted in a NERI proposal dealing with turbo-machinery design being approved for funding beginning in FY01. Two other NERI proposals, dealing with the development of a burnup “meter” and modularization techniques, were also funded in which the MIT team will be a participant. · A South African MPBR fuel testing proposal is pending ($7.0M over nine years).

Petti, David Andrew

2000-07-01T23:59:59.000Z

160

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...

Note: This page contains sample records for the topic "research reactor spent" 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

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

162

FUEL ELEMENTS FOR THE ARGONNE ADVANCED RESEARCH REACTOR  

SciTech Connect

The core design and the fuel element concept for the high-flux Argonne Advanced Research Reactor are presented. The core is cooled and moderated by light water and utilizes beryllium as a reflector. The fuel element assembly is rhomboidal in cross section and consists of 27 plates fastened together at their edges by dovetailed locking keys, and at each end by end fittings. Each fuel plate is 40 mils thick and contains a uniform dispersion of highly enriched UO/ sub 2/ particles, up to a maximum of 37 wt%, in a matrix of sintered stainless steel powder. A 5 mil thick stainless steel cladding is metallurgically bonded to each side of the fueled matrix. (N.W.R.)

Adolph, N.R.; Silberstein, M.S.; Weinstein, A.

1962-01-01T23:59:59.000Z

163

MCNP-model for the OAEP Thai Research Reactor  

SciTech Connect

An MCNP input was prepared for the Thai Research Reactor, making extensive use of the MCNP geometry`s lattice feature that allows a flexible and easy rearrangement of the core components and the adjustment of the control elements. The geometry was checked for overdefined or undefined zones by two-dimensional plots of cuts through the core configuration with the MCNP geometry plotting capabilities, and by a three-dimensional view of the core configuration with the SABRINA code. Cross sections were defined for a hypothetical core of 67 standard fuel elements and 38 low-enriched uranium fuel elements--all filled with fresh fuel. Three test calculations were performed with the MCNP4B-code to obtain the multiplication factor for the cases with control elements fully inserted, fully withdrawn, and at a working position.

Gallmeier, F.X.; Tang, J.S.; Primm, R.T. III

1998-06-01T23:59:59.000Z

164

Sodium fast reactor fuels and materials : research needs.  

SciTech Connect

An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

Denman, Matthew R.; Porter, Douglas (Idaho National Laboratory, Idaho Falls, ID); Wright, Art (Argonne National Laboratory Argonne, IL); Lambert, John (Argonne National Laboratory Argonne, IL); Hayes, Steven (Idaho National Laboratory, Idaho Falls, ID); Natesan, Ken (Argonne National Laboratory Argonne, IL); Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Garner, Frank (Radiation Effects Consulting. Richland, WA); Walters, Leon (Advanced Reactor Concepts, Idaho Falls, ID); Yacout, Abdellatif (Argonne National Laboratory Argonne, IL)

2011-09-01T23:59:59.000Z

165

High uranium density dispersion fuel for the reduced enrichment of research and test reactors program.  

E-Print Network (OSTI)

??This work describes the fabrication of a high uranium density fuel for the Reduced Enrichment of Research and Test Reactors Program. In an effort to… (more)

[No author

2006-01-01T23:59:59.000Z

166

Final Site-Specific Decommissioning Inspection Report for the University of Washington Research and Test Reactor  

SciTech Connect

Report of site-specific decommissioning in-process inspection activities at the University of Washington Research and Test Reactor Facility.

Sarah Roberts

2006-10-18T23:59:59.000Z

167

The inverse kinetics method and PID compensation of the Annular Core Research Reactor.  

E-Print Network (OSTI)

??This thesis explores the development of a model describing the Annular Core Research Reactor (ACRR), the application of the inverse kinetics method to calculate the… (more)

Garnas, Benjamin

2008-01-01T23:59:59.000Z

168

Sodium fast reactor safety and licensing research plan. Volume I.  

SciTech Connect

This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

Sofu, Tanju (Argonne National Laboratory, Argonne, IL); LaChance, Jeffrey L.; Bari, R. (Brokhaven National Laboratory Upton, NY); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.; Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN)

2012-05-01T23:59:59.000Z

169

The initial results of research on two-step cascades in the Dalat research reactor  

E-Print Network (OSTI)

By the financial support of Vietnam Atomic Energy Commission and kind cooperation of Frank Laboratory, in the year of 2005 a measure system based on summation of amplitude pulses was established on the tangential channel of Dalat Research Reactor. After a serial of testing, the measure system was explored. In this, we would like to show the initial results were gotten with 36-Cl isotope.

Nguyen Xuan Hai; Pham Dinh Khang; Vuong Huu Tan; Ho Huu Thang; A. M. Sukhovoj; V. A. Khitrov

2013-07-11T23:59:59.000Z

170

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

171

Development and transfer of fuel fabrication and utilization technology for research reactors  

SciTech Connect

Approximately 300 research reactors supplied with US-enriched uranium are currently in operation in about 40 countries, with a variety of types, sizes, experiment capabilities and applications. Despite the usefulness and popularity of research reactors, relatively few innovations in their core design have been made in the last fifteen years. The main reason can be better understood by reviewing briefly the history of research reactor fuel technology and enrichment levels. Stringent requirements on the enrichment of the uranium to be used in research reactors were considered and a program was launched to assist research reactors in continuing their operation with the new requirements and with minimum penalties. The goal of the new program, the Reduced Enrichment Research and Test Reactor (RERTR) Program, is to develop the technical means to utilize LEU instead of HEU in research reactors without significant penalties in experiment performance, operating costs, reactor modifications, and safety characteristics. This paper reviews briefly the RERTR Program activities with special emphasis on the technology transfer aspects of interest to this conference.

Travelli, A.; Domagala, R.F.; Matos, J.E.; Snelgrove, J.L.

1982-01-01T23:59:59.000Z

172

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

173

West Valley facility spent fuel handling, storage, and shipping experience  

Science Conference Proceedings (OSTI)

The result of a study on handling and shipping experience with spent fuel are described in this report. The study was performed by Pacific Northwest Laboratory (PNL) and was jointly sponsored by the US Department of Energy (DOE) and the Electric Power Research Institute (EPRI). The purpose of the study was to document the experience with handling and shipping of relatively old light-water reactor (LWR) fuel that has been in pool storage at the West Valley facility, which is at the Western New York Nuclear Service Center at West Valley, New York and operated by DOE. A subject of particular interest in the study was the behavior of corrosion product deposits (i.e., crud) deposits on spent LWR fuel after long-term pool storage; some evidence of crud loosening has been observed with fuel that was stored for extended periods at the West Valley facility and at other sites. Conclusions associated with the experience to date with old spent fuel that has been stored at the West Valley facility are presented. The conclusions are drawn from these subject areas: a general overview of the West Valley experience, handling of spent fuel, storing of spent fuel, rod consolidation, shipping of spent fuel, crud loosening, and visual inspection. A list of recommendations is provided. 61 refs., 4 figs., 5 tabs.

Bailey, W.J.

1990-11-01T23:59:59.000Z

174

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

175

U.S. Department of Energy Program of International Technical Cooperation for Research Reactor Utilization  

SciTech Connect

The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; identification of competition and estimation of potential for market penetration; integration of technical constraints; estimation of cash flow streams; and case studies.

Chong, D.; Manning, M.; Ellis, R.; Apt, K.; Flaim, S.; Sylvester, K.

2004-10-03T23:59:59.000Z

176

Nuclear reactor and materials science research: Technical report, May 1, 1985-September 30, 1986  

Science Conference Proceedings (OSTI)

Throughout the 17-month period of its grant, May 1, 1985-September 30, 1986, the MIT Research Reactor (MITR-II) was operated in support of research and academic programs in the physical and life sciences and in related engineering fields. The reactor was operated 4115 hours during FY 1986 and for 6080 hours during the entire 17-month period, an average of 82 hours per week. Utilization of the reactor during that period may be classified as follows: neutron beam tube research; nuclear materials research and development; radiochemistry and trace analysis; nuclear medicine; radiation health physics; computer control of reactors; dose reduction in nuclear power reactors; reactor irradiations and services for groups outside MIT; MIT Research Reactor. Data on the above utilization for FY 1986 show that the MIT Nuclear Reactor Laboratory (NRL) engaged in joint activities with nine academic departments and interdepartmental laboratories at MIT, the Charles Stark Draper Laboratory in Cambridge, and 22 other universities and nonprofit research institutions, such as teaching hospitals.

Not Available

1987-05-11T23:59:59.000Z

177

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

178

Monitoring and Control Research Using a University Reactor and SBWR Test-Loop  

Science Conference Proceedings (OSTI)

The existing hybrid simulation capability of the Penn State Breazeale nuclear reactor was expanded to conduct research for monitoring, operations and control. Hybrid simulation in this context refers to the use of the physical time response of the research reactor as an input signal to a real-time simulation of power-reactor thermal-hydraulics which in-turn provides a feedback signal to the reactor through positioning of an experimental changeable reactivity device. An ECRD is an aluminum tube containing an absorber material that is positioned in the central themble of the reactor kinetics were used to expand the hybrid reactor simulation (HRS) capability to include out-of-phase stability characteristics observed in operating BWRs.

Robert M. Edwards

2003-09-28T23:59:59.000Z

179

Gas release driven dynamics in research reactors piping  

SciTech Connect

Analysis of the physical and chemical processes of radiolysis gas production, air absorption, diffusion controlled gas release and transport in the coolant cleaning system of the research reactor FRM II, which is now being in routine power operation in Munich, Germany, lead to the following conclusions: 1) The steady state pressure distribution in the siphon pipe allows that the horizontal part of the siphon pipe is filled with air. The air is isolated by about 1 m water column from the main pipe of the coolant cleaning system (CCS). This is a stable steady state. It has two positive impacts on the normal operation of the CCS: (a) there is effectively no bypass flow; (b) The air can not be transported through the pipe and therefore no deterioration of the pump performance is expected from the function of the siphon pipe. 2) Radiolysis gas production for coolant, that initially does not contain dissolved air, does not lead to any problem for the system. The gases are dissolved in the coolant at 2.2 bar and are not released for pressures reduction to about 1 bar, which is the minimum pressure in the CCS. 3) Assuming hypothetically a radiolysis gas production for coolant, which initially does contain dissolved air close to its saturation, leads to gas slug formation and its transport up to the pump. This could reduce the pump head and could lead to distortion of the normal operation. Systematic measurement of the hydrogen in the primary system at 100% power indicated, that this state is not realized in the system. The observed H{sub 2} concentration was between 0.016 e-6 and 0.380 e-6 which is of no concern at all. (authors)

Kolev, Nikolay Ivanov; Roloff-Bock, Iris; Schlicht, Gerhard [Framatome ANP, P.O. Box 3220, D-91058, Erlangen (Germany)

2006-07-01T23:59:59.000Z

180

Spent fuel transportation in the United States: commercial spent fuel shipments through December 1984  

Science Conference Proceedings (OSTI)

This report has been prepared to provide updated transportation information on light water reactor (LWR) spent fuel in the United States. Historical data are presented on the quantities of spent fuel shipped from individual reactors on an annual basis and their shipping destinations. Specifically, a tabulation is provided for each present-fuel shipment that lists utility and plant of origin, destination and number of spent-fuel assemblies shipped. For all annual shipping campaigns between 1980 and 1984, the actual numbers of spent-fuel shipments are defined. The shipments are tabulated by year, and the mode of shipment and the casks utilized in shipment are included. The data consist of the current spent-fuel inventories at each of the operating reactors as of December 31, 1984. This report presents historical data on all commercial spent-fuel transportation shipments have occurred in the United States through December 31, 1984.

Not Available

1986-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" 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

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

182

Risk management at the Oak Ridge National Laboratory research reactors  

SciTech Connect

In November of 1986, the High Flux Isotope Reactor (HFIR) was shut down by Oak Ridge National Laboratory (ORNL) due to a concern regarding embrittlement of the reactor vessel. A massive review effort was undertaken by ORNL and the Department of Energy (DOE). This review resulted in an extensive list of analyses and design modifications to be completed before restart could take place. The review also focused on the improvement of management practices including implementation of several of the Institute of Nuclear Power Operations (INPO) requirements. One of the early items identified was the need to perform a Probabilistic Risk Assessment (PRA) on the reactor. It was decided by ORNL management that this PRA would not be just an exercise to assess the ``bottom`` line in order to restart, but would be used to improve the overall safety of the reactor, especially since resources (both manpower and dollars) were severely limited. The PRA would become a basic safety tool to be used instead of a more standard deterministic approach to safety used in commercial reactor power plants. This approach was further reinforced, because the reactor was nearly 25 years old at this time, and the design standards and regulations had changed significantly since the original design, and many of the safety issues could not be addressed by compliance to codes and standards.

Flanagan, G.F.; Linn, M.A.; Proctor, L.D.; Cook, D.H.

1994-12-31T23:59:59.000Z

183

Reactor Safety Research Programs Quarterly Report July - September 1981  

SciTech Connect

This document summarizes the work performed by Pacific Northwest laboratory (PNL) from July 1 through September 30, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR} steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Edler, S. K.

1982-01-01T23:59:59.000Z

184

USA NRC/RSR Data Bank System and Reactor Safety Research Data Repository (RSRDR)  

SciTech Connect

The United States Nuclear Regulatory Commission (NRC), through its Division of Reactor Safety Research (RSR) of the Office of Nuclear Regulatory Research, has established the NRC/RSR Data Bank Program to collect, process, and make available data from the many domestic and foreign water reactor safety research programs. An increasing number of requests for data and/or calculations generated by NRC Contractors led to the initiation of the program which allows timely and direct access to water reactor safety data in a manner most useful to the user. The program consists of three main elements: data sources, service organizations, and a data repository.

Maskewitz, B.F.; Bankert, S.F.

1979-01-01T23:59:59.000Z

185

Thermal-hydraulic aspects of flow inversion in a research reactor  

SciTech Connect

PARET, a neutronics and thermal-hydraulics computer code, has been modified to account for natural convection in a reactor core. The code was then used to analyze the flow inversion that occurs in a reactor with heat removal by forced convection in the downward direction after a pump failure. Typical results are shown for a number of parameters. Research reactors normally operating much above ten MW are predicted to experience nucleate boiling in the event of a flow inversion. Comparison with experimental results from the Belgian BR2 reactor indicated general agreement although nucleate boiling that was analytically predicted was not noted in the BR2 data.

Smith, R.S.; Woodruff, W.L.

1986-11-01T23:59:59.000Z

186

Light Water Reactor Fuel Cladding Research and Testing | ornl.gov  

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

Light Water Reactor Fuel Cladding Research Light Water Reactor Fuel Cladding Research June 01, 2013 Severe Accident Test Station ORNL is the focus point for Light Water Reactor (LWR) fuel cladding research and testing. The purpose of this research is to furnish U.S. industry (EPRI, Areva, Westinghouse), and regulators (NRC) with much-needed data supporting safe and economical nuclear power generation and used fuel management. LWR fuel cladding work is tightly integrated with ORNL accident tolerant fuel development and used fuel disposition programs thereby providing a powerful capability that couples basic materials science research with the nuclear applications research and development. The ORNL LWR fuel cladding program consists of five complementary areas of research: Accident tolerant fuel and cladding material testing under design

187

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program  

SciTech Connect

The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

2012-09-01T23:59:59.000Z

188

Two U.S. University Research Reactors to be Converted From Highly Enriched  

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

U.S. University Research Reactors to be Converted From Highly U.S. University Research Reactors to be Converted From Highly Enriched Uranium to Low-Enriched Uranium Two U.S. University Research Reactors to be Converted From Highly Enriched Uranium to Low-Enriched Uranium April 11, 2005 - 11:34am Addthis WASHINGTON, D.C. - As part of the Bush administration's aggressive effort to reduce the amount of weapons-grade nuclear material worldwide, Secretary of Energy Samuel W. Bodman announced today that the Department of Energy (DOE) has begun to convert research reactors from using highly-enriched uranium (HEU) to low-enriched uranium fuel (LEU) at the University of Florida and Texas A&M University. This effort, by DOE's National Nuclear Security Administration (NNSA) and the Office of Nuclear Energy, Science and Technology, are the latest steps

189

Effect of reduced enrichment on the fuel cycle for research reactors  

SciTech Connect

The new fuels developed by the RERTR Program and by other international programs for application in research reactors with reduced uranium enrichment (<20% EU) are discussed. It is shown that these fuels, combined with proper fuel-element design and fuel-management strategies, can provide at least the same core residence time as high-enrichment fuels in current use, and can frequently significantly extend it. The effect of enrichment reduction on other components of the research reactor fuel cycle, such as uranium and enrichment requirements, fuel fabrication, fuel shipment, and reprocessing are also briefly discussed with their economic implications. From a systematic comparison of HEU and LEU cores for the same reference research reactor, it is concluded that the new fuels have a potential for reducing the research reactor fuel cycle costs while reducing, at the same time, the uranium enrichment of the fuel.

Travelli, A.

1982-01-01T23:59:59.000Z

190

Demonstration of the reactivity constraint approach on SNL's annual core research reactor  

Science Conference Proceedings (OSTI)

This paper reports on the initial demonstration of the reactivity constraint approach and its implementing algorithm, the MIT-CSDL Non-Linear Digital Controller, on the annual core research reactor (ACCR) that is operated by the Sandia National Laboratories. This demonstration constituted the first use of reactivity constraints for the closed-loop, digital control of reactor power on a facility other than the Massachusetts Institute of Technology's (MIT's) research reactor (MITR-II). Also, because the ACRR and the MITR-II are of very different design, these trials established the generic nature of the reactivity constraint approach.

Bernard, J.A.; Kwok, K.S.; Wyant, F.J.; Thome, F.V.

1989-01-01T23:59:59.000Z

191

A Brief History i-l Research Reactors  

E-Print Network (OSTI)

stainless steel sam- ples in the High Flux Isotope Reactor (HFIR) at tem- peratures of 380 to 680" with up/cm' to balance the gas pressure were used m their calculation. A comparison of the results with HFIR and the HFIR ex- perimental data is presented in section 5. Applications of the model to various fusion designs

192

In accounts of seminal neutron research at ORNL's Graphite Reactor,  

E-Print Network (OSTI)

requested permission to set up an X-ray diffractometer he had brought from the University of Chicago for his of the Graphite Reactor. "I was a student at the University of Chicago in 1942 when Enrico Fermi was doing his. 25 Scrooge (OR Playhouse) Nov. 27 Football: UT vs. Kentucky Dec. 4 Fiddler on the Roof Dec. 11 Best

193

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

194

Metals removal from spent salts  

DOE Patents (OSTI)

A method and apparatus for removing metal contaminants from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents may be added to precipitate the metal oxide and/or the metal as either metal oxide, metal hydroxide, or as a salt. The precipitated materials are filtered, dried and packaged for disposal as waste or can be immobilized as ceramic pellets. More than about 90% of the metals and mineral residues (ashes) present are removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be spray-dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 1.0 ppm of contaminants.

Hsu, Peter C. (Pleasanton, CA); Von Holtz, Erica H. (Livermore, CA); Hipple, David L. (Livermore, CA); Summers, Leslie J. (Livermore, CA); Brummond, William A. (Livermore, CA); Adamson, Martyn G. (Danville, CA)

2002-01-01T23:59:59.000Z

195

Actinide removal from spent salts  

DOE Patents (OSTI)

A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 0.1 ppm of thorium or uranium.

Hsu, Peter C. (Pleasanton, CA); von Holtz, Erica H. (Livermore, CA); Hipple, David L. (Livermore, CA); Summers, Leslie J. (Livermore, CA); Adamson, Martyn G. (Danville, CA)

2002-01-01T23:59:59.000Z

196

Materials Research Needs for Near-Term Nuclear Reactors  

Science Conference Proceedings (OSTI)

Technical Paper / NSF Workshop on the Research Needs of the Next Generation Nuclear Power Technology / Material

John R. Weeks

197

Thermal hydraulic limits analysis for the MIT Research Reactor low enrichment uranium core conversion using statistical propagation of parametric uncertainties  

E-Print Network (OSTI)

The MIT Research Reactor (MITR) is evaluating the conversion from highly enriched uranium (HEU) to low enrichment uranium (LEU) fuel. In addition to the fuel element re-design from 15 to 18 plates per element, a reactor ...

Chiang, Keng-Yen

2012-01-01T23:59:59.000Z

198

Decommissioning Anddismantling Of The Research Reactor Salaspils. I. Conceptual Study And The First Results.  

E-Print Network (OSTI)

In May 1995, the Latvian government decided to shut down the Research Reactor Salaspils (SRR) and to dis pense with nuclear energy in future. The reactor is out of operation since July 1998. A conceptual study for the decommissioning of SRR has been carried out by Noell-KRC-Energie- und Umweltlechnik GmbH at 1998-1999 years. The Latvian government decided in October 26 1999 to start the direct dismantling to "green field" at 2001 year. The first decommissioning and dismantling results from preparation measures in 1999 year are presented and discussed. The main efforts was devoted to collecting and conditioning of "historical" radioactive wastes from different storages outside and inside of reactor hall. All non-radioactive equipments and materials outside of reactor buildings were free-released and dismantled for reusing and conventional disposing. Weakly contaminated materials from reactor hall were collected and removed for free-release measurements. 1.0 INTRODUCTION The res...

Andris Abramenkovs Ministry; Andris Abramenkovs; Arnis Ezergailis; State Enterprise “vides Projekti; Dzintars Kalninš

2000-01-01T23:59:59.000Z

199

Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report  

Science Conference Proceedings (OSTI)

This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOE’s Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

2002-11-01T23:59:59.000Z

200

Management of HFIR spent fuel  

Science Conference Proceedings (OSTI)

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) has been unable to ship its spent fuel off-site for reprocessing since 1985. The HFIR storage pools are expected to fill up by the end of 1994. If a management alternative to existing HFIR pool storage is not identified and implemented by that time, the HFIR will be forced to shut down. This study identified and investigated five alternatives to managing the HFIR spent fuel, to determine the feasibility of implementing each in time to prevent shutdown of the HFIR: (1) increasing HFIR pool storage capacity, (2) storing the spent fuel at another ORNL pool, (3) storing the spent fuel in one or more hot cells at ORNL, (4) shipping the spent fuel off-site for reprocessing or storage elsewhere, and (5) installing a dedicated dry storage facility at ORNL. Of the alternatives investigated, only two could prevent the shutdown of the HFIR in the near term: increasing HFIR pool storage capacity or shipping the spent fuel off-site. Both options have been vigorously pursued because neither is assured of success, and at least one of the options must be successfully implemented if the HFIR is to continue operation. In addition, a third option was selected for implementation as an intermediate-term storage solution: installing a dedicated dry storage facility for the HFIR. An intermediate-term storage solution is needed because neither of the short-term solutions could ensure long-term continued operation of the HFIR.

Green, V.M.; Begovich, J.M.; Flanagan, G.F. [Oak Ridge National Lab., TN (United States); Lotts, A.L.

1994-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" 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

MITR-III: Upgrade and relicensing studies for the MIT Research Reactor. Second annual report  

SciTech Connect

The current operating license of the MIT research reactor will expire on May 7, 1996 or possibly a few years later if the US Nuclear Regulatory Commission agrees that the license period can start with the date of initial reactor operation. Driven by the imminent expiration of the operating license, a team of nuclear engineering staff and students have begun a study of the future options for the MIT Research Reactor. These options have included the range from a major rebuilding of the reactor to its decommissioning. This document reports the results of a two year intensive activity which has been supported by a $148,000 grant from the USDOE contract Number DEFG0293ER75859, approximately $100,000 of internal MIT funds and Nuclear Engineering Department graduate student fellowships as well as assistance from international visiting scientists and engineers.

Trosman, H.G. [ed.; Lanning, D.D.; Harling, O.K.

1994-08-01T23:59:59.000Z

202

Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel  

SciTech Connect

The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. Results of design and safety analyses performed by the RERTR Program at the Argonne National Laboratory for LEU conversion of the GTRR are summarized. Only those parameters which could change as a result of replacing the fuel are addressed. The performance of the reactor and all safety margins with LEU fuel are expected to be about the same as those with the current HEU fuel.

Matos, J.E.; Mo, S.C.; Woodruff, W.L.

1992-01-01T23:59:59.000Z

203

Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel  

Science Conference Proceedings (OSTI)

The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. Results of design and safety analyses performed by the RERTR Program at the Argonne National Laboratory for LEU conversion of the GTRR are summarized. Only those parameters which could change as a result of replacing the fuel are addressed. The performance of the reactor and all safety margins with LEU fuel are expected to be about the same as those with the current HEU fuel.

Matos, J.E.; Mo, S.C.; Woodruff, W.L.

1992-12-31T23:59:59.000Z

204

Leu conversion status of U.S. research reactors: September 1996  

SciTech Connect

At the request of the Department of Energy, the RERTR Program has summarized the conversion status of research and test reactors in the United States and has made estimates of the uranium densities that would be needed to convert the reactors with power levels greater than or equal to 1 MW from Highly Enriched Uranium (HEU) (greater than or equal to 20% U-235) to Lightly Enriched Uranium (LEU) (less than 20% U-235) fuels. Detailed conversion studies for each of the reactors need to be completed in order to establish the feasibility of using LEU fuels.

Matos, J.E.

1996-09-01T23:59:59.000Z

205

Education program at the Massachusetts Institute of Technology research reactor for pre-college science teachers  

Science Conference Proceedings (OSTI)

A Pre-College Science Teacher (PCST) Seminar program has been in place at the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory for 4 yr. The purpose of the PCST program is to educate teachers in nuclear technology and to show teachers, and through them the community, the types of activities performed at research reactors. This paper describes the background, content, and results of the MIT PCST program.

Hopkins, G.R.; Fecych, W.; Harling, O.K.

1989-01-01T23:59:59.000Z

206

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

207

A case study on the influence of THM coupling on the near field safety of a spent fuel repository in sparsely fractured granite  

E-Print Network (OSTI)

geological disposal of spent CANDU fuel in Canada, a safetyhypothetical repository for spent CANDU fuel in the Canadianbuffer. The waste form: CANDU reactors in Canada are fuelled

Nguyen, T.S.

2009-01-01T23:59:59.000Z

208

Using low-enriched uranium in research reactors: The RERTR program  

SciTech Connect

The goal of the RERTR program is to minimize and eventually eliminate use of highway enriched uranium (HEU) in research and test reactors. The program has been very successful, and has developed low-enriched uranium (LEU) fuel materials and designs which can be used effectively in approximately 90 percent of the research and test reactors which used HEU when the program began. This progress would not have been possible without active international cooperation among fuel developers, commercial vendors, and reactor operators. The new tasks which the RERTR program is undertaking at this time include development of new and better fuels that will allow use of LEU fuels in all research and test reactors; cooperation with Russian laboratories, which will make it possible to minimize and eventually eliminate use of HEU in research reactors throughout the world, irrespective of its origin; and development of an LEU-based process for the production of {sup 99}Mo. Continuation and intensification of international cooperation are essential to the achievement of the ultimate goals of the RERTR program.

Travelli, A.

1994-05-01T23:59:59.000Z

209

Major Safety Aspects of Advanced Candu Reactor and Associated Research and Development  

Science Conference Proceedings (OSTI)

The Advanced Candu{sup R} Reactor design is built on the proven technology of existing Candu plants and on AECL's knowledge base acquired over decades of nuclear power plant design, engineering, construction and research. Two prime objectives of ACR-700TM1 are cost reduction and enhanced safety. To achieve them some new features were introduced and others were improved from the previous Candu 6 and Candu 9 designs. The ACR-700 reactor design is based on the modular concept of horizontal fuel channels surrounded by a heavy water moderator, the same as with all Candu reactors. The major novelty in the ACR-700 is the use of slightly enriched fuel and light water as coolant circulating in the fuel channels. This results in a more compact reactor design and a reduction of heavy water inventory, both contributing to a significant decrease in cost compared to Candu reactors, which employ natural uranium as fuel and heavy water as coolant. The reactor core design adopted for ACR-700 also has some features that have a bearing on inherent safety, such as negative power and coolant void reactivity coefficient. Several improvements in engineered safety have been made as well, such as enhanced separation of the safety support systems. Since the ACR-700 design is an evolutionary development of the currently operating Candu plants, limited research is required to extend the validation database for the design and the supporting safety analysis. A program of safety related research and development has been initiated to address the areas where the ACR-700 design is significantly different from the Candu designs. This paper describes the major safety aspects of the ACR-700 with a particular focus on novel features and improvements over the existing Candu reactors. It also outlines the key areas where research and development efforts are undertaken to demonstrate the effectiveness and robustness of the design. (authors)

Bonechi, M.; Wren, D.J.; Hopwood, J.M. [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

2002-07-01T23:59:59.000Z

210

Thermal hydraulics analysis of the MIT research reactor in support of a low enrichment uranium (LEU) core conversion.  

E-Print Network (OSTI)

??The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density… (more)

Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

2008-01-01T23:59:59.000Z

211

Evaluation of the thermal-hydraulic operating limits of the HEU-LEU transition cores for the MIT Research Reactor.  

E-Print Network (OSTI)

??The MIT Research Reactor (MITR) is in the process of conducting a design study to convert from High Enrichment Uranium (HEU) fuel to Low Enrichment… (more)

Wang, Yunzhi (Yunzhi Diana)

2009-01-01T23:59:59.000Z

212

N-Reactor Department research and development budget for FY-1967 and revision of budget for FY-1966  

SciTech Connect

This report provides the N-Reactor research and development budget for fiscal year 1967 and modifications of the budget for fiscal year 1966.

1965-04-19T23:59:59.000Z

213

Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors  

SciTech Connect

The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

1993-07-01T23:59:59.000Z

214

A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor.  

SciTech Connect

The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% {sup 235}U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm{sup 3} and 3.8 gU/cm{sup 3} are required to match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively.

Bretscher, M. M.

1998-10-14T23:59:59.000Z

215

Comprehensive Thermal Hydraulics Research of the Very High Temperature Gas Cooled Reactor  

SciTech Connect

The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; David Petti; Hyung Kang

2010-10-01T23:59:59.000Z

216

Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor  

SciTech Connect

A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams.

Hu Linwen; Bernard, John A.; Hejzlar, Pavel; Kohse, Gordon [Massachusetts Institute of Technology (United States)

2005-05-15T23:59:59.000Z

217

Revised ROD for FEIS on Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel (  

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

720 720 Federal Register / Vol. 61, No. 144 / Thursday, July 25, 1996 / Notices 1995 (44 U. S. C. Chapter 35) requires that the Office of Management and Budget (OMB) provide interested Federal agencies and the public an early opportunity to comment on information collection requests. OMB may amend or waive the requirement for public consultation to the extent that public participation in the approval process would defeat the purpose of the information collection, violate State or Federal law, or substantially interfere with any agency's ability to perform its statutory obligations. The Director of the Information Resources Group publishes this notice containing proposed information collection requests prior to submission of these requests to OMB. Each proposed information collection,

218

Spent fuel integrity during transportation  

SciTech Connect

The conditions of recent shipments of light water reactor spent fuel were surveyed. The radioactivity level of cask coolant was examined in an attempt to find the effects of transportation on LWR fuel assemblies. Discussion included potential cladding integrity loss mechanisms, canning requirements, changes of radioactivity levels, and comparison of transportation in wet or dry media. Although integrity loss or degradation has not been identified, radioactivity levels usually increase during transportation, especially for leaking assemblies.

Funk, C.W.; Jacobson, L.D.

1980-01-01T23:59:59.000Z

219

Spent Fuel Pool Accident Characteristics  

Science Conference Proceedings (OSTI)

Spent fuel pools (SFPs) at nuclear reactor sites contain used fuel assemblies, control rods, used radioactive sources, and used instrumentation. Cooling of the used fuel is required to remove the decay heat generated by radioactive decay.BackgroundThe SFPs include heat removal systems to provide methods to cool the used fuel and inventory makeup systems as backup methods to preserve water inventory if the SFP cooling system is ineffective. These two methods ...

2013-05-27T23:59:59.000Z

220

Foreign research reactor irradiated nuclear fuel inventories containing HEU and LEU of United States origin  

SciTech Connect

This report provides estimates of foreign research reactor inventories of aluminum-based and TRIGA irradiated nuclear fuel elements containing highly enriched and low enriched uranium of United States origin that are anticipated in January 1996, January 2001, and January 2006. These fuels from 104 research reactors in 41 countries are the same aluminum-based and TRIGA fuels that were eligible for receipt under the Department of Energy`s Offsite Fuels Policy that was in effect in 1988. All fuel inventory and reactor data that were available as of December 1, 1994, have been included in the estimates of approximately 14,300 irradiated fuel elements in January 1996, 18,800 in January 2001, and 22,700 in January 2006.

Matos, J.E.

1994-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" 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.


221

DESTRUCTIVE EXAMINATION OF 3-CYCLE LWR (LIGHT WATER REACTOR) FUEL RODS FROM TURKEY POINT UNIT 3 FOR THE CLIMAX - SPENT FUEL TEST  

DOE Green Energy (OSTI)

The destructive examination results of five light water reactor rods from the Turkey Point Unit 3 reactor are presented. The examinations included fission gas collection and analyses, burnup and hydrogen analyses, and a metallographic evaluation of the fuel, cladding, oxide, and hydrides. The rods exhibited a low fission gas release with all other results appearing representative for pressurized water reator fuel rods with similar burnups (28 GWd/MTU) and operating histories.

ATKIN SD

1981-06-01T23:59:59.000Z

222

MYRRHA a multi-purpose hybrid research reactor for high-tech applications  

SciTech Connect

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is the flexible experimental accelerator driven system (ADS) in development at SCK-CEN. MYRRHA is able to work both in subcritical (ADS) as in critical mode. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for generation IV (GEN IV) systems, material developments for fusion reactors, radioisotope production and industrial applications, such as Si-doping. MYRRHA will also demonstrate the ADS full concept by coupling the three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow the study of efficient transmutation of high-level nuclear waste. MYRRHA is based on the heavy liquid metal technology and so it will contribute to the development of lead fast reactor (LFR) technology and in critical mode, MYRRHA will play the role of European technology pilot plant in the roadmap for LFR. In this paper the historical evolution of MYRRHA and the rationale behind the design choices is presented and the latest configuration of the reactor core and primary system is described. (authors)

Abderrahim, H. A.; Baeten, P. [SCK CEN, Boeretang 200, 2400 Mol (Belgium)

2012-07-01T23:59:59.000Z

223

Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel  

SciTech Connect

This document presents information concerning: analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU; changes to technical specifications mandated by the conversion of the GTRR to low enrichment fuel; changes in the Safety Analysis Report mandated by the conversion of the GTRR to low enrichment fuel; and copies of all changed pages of the SAR and the technical specifications.

Matos, J.E.; Mo, S.C.; Woodruff, W.L.

1992-09-01T23:59:59.000Z

224

Prototype spent-fuel canister design, analysis, and test  

SciTech Connect

Sandia National Laboratories was asked by the US Energy Research and Development Administration (now US Department of Energy) to design the spent fuel shipping cask system for the Clinch River Breeder Reactor Plant (CRBRP). As a part of this task, a canister which holds liquid sodium and the spent fuel assembly was designed, analyzed, and tested. The canister body survived the regulatory Type-B 9.1-m (30-ft) drop test with no apparent leakage. However, the commercially available metal seal used in this design leaked after the tests. This report describes the design approach, analysis, and prototype canister testing. Recommended work for completing the design, when funding is available, is included.

Leisher, W.B.; Eakes, R.G.; Duffey, T.A.

1982-03-01T23:59:59.000Z

225

Monte Carlo simulation of the Massachusetts Institute of Technology Research Reactor  

SciTech Connect

The three-dimensional continuous-energy MCNP Monte Carlo code is used to develop a versatile and accurate reactor physics model of the Massachusetts Institute of Technology Research Reactor 2 (MITR-2). The validation of the model against existing experimental data is presented. Core multiplication factors as well as fast neutron in-core flux measurements were used in the validation process. The agreement between the MCNP predictions and the experimentally determined values is very good, which indicates that the Monte Carlo model is correctly simulating the MITR-2.

Redmond, E.L. II; Yanch, J.C.; Harling, O.K. (Massachusetts Inst. of Tech., Cambridge, MA (United States). Nuclear Engineering Dept.)

1994-04-01T23:59:59.000Z

226

Neutronic safety parameters and transient analyses for potential LEU conversion of the Budapest Research Reactor.  

SciTech Connect

An initial safety study for potential LEU conversion of the Budapest Research Reactor was completed. The study compares safety parameters and example transients for reactor cores with HEU and LEU fuels. Reactivity coefficients, kinetic parameters and control rod worths were calculated for cores with HEU(36%) UAl alloy fuel and UO{sub 2}-Al dispersion fuel, and with LEU (19.75%)UO{sub 2}-Al dispersion fuel that has a uranium density of about 2.5 g/cm{sup 3}. A preliminary fuel conversion plan was developed for transition cores that would convert the BRR from HEU to LEU fuel after the process is begun.

Pond, R. B.; Hanan, N. A.; Matos, J. E.; Maraczy, C.

1999-09-27T23:59:59.000Z

227

Strategic Plan for Light Water Reactor Research and Development  

Science Conference Proceedings (OSTI)

The purpose of this strategic plan is to establish a framework that will allow the Department of Energy (DOE) and the nuclear power industry to jointly plan the nuclear energy research and development (R&D) agenda important to achieving the Nation's energy goals. This strategic plan has been developed to focus on only those R&D areas that will benefit from a coordinated government/industry effort. Specifically, this plan focuses on safely sustaining and expanding the electricity output from currently operating nuclear power plants and expanding nuclear capacity through the deployment of new plants. By focusing on R&D that addresses the needs of both current and future nuclear plants, DOE and industry will be able to take advantage of the synergism between these two technology areas, thus improving coordination, enhancing efficiency, and further leveraging public and private sector resources. By working together under the framework of this strategic plan, DOE and the nuclear industry reinforce their joint commitment to the future use of nuclear power and the National Energy Policy's goal of expanding its use in the United States. The undersigned believe that a public-private partnership approach is the most efficient and effective way to develop and transfer new technologies to the marketplace to achieve this goal. This Strategic Plan is intended to be a living document that will be updated annually.

None

2004-02-01T23:59:59.000Z

228

Radiological survey support activities for the decommissioning of the Ames Laboratory Research Reactor Facility, Ames, Iowa  

SciTech Connect

At the request of the Engineering Support Division of the US Department of Energy-Chicago Operations Office and in accordance with the programmatic overview/certification responsibilities of the Department of Energy Environmental and Safety Engineering Division, the Argonne National Laboratory Radiological Survey Group conducted a series of radiological measurements and tests at the Ames Laboratory Research Reactor located in Ames, Iowa. These measurements and tests were conducted during 1980 and 1981 while the reactor building was being decontaminated and decommissioned for the purpose of returning the building to general use. The results of these evaluations are included in this report. Although the surface contamination within the reactor building could presumably be reduced to negligible levels, the potential for airborne contamination from tritiated water vapor remains. This vapor emmanates from contamination within the concrete of the building and should be monitored until such time as it is reduced to background levels. 2 references, 8 figures, 6 tables.

Wynveen, R.A.; Smith, W.H.; Sholeen, C.M.; Justus, A.L.; Flynn, K.F.

1984-09-01T23:59:59.000Z

229

STORED ENERGY: GROWTH AND ANNEALING STATUS OF GRAPHITE MODERATOR IN THE BNL RESEARCH REACTOR. Final Report  

SciTech Connect

The present sthtus, past annealing procedures and experiences, future annealing procedures, annealing sehedule, revised annealing procedure (1958), procedure for combating a graphite fire in fuel channel, high-temperature stored energy, and graphite burning experiments are reportcd for the BNL Research Reactor. The following subjccts are discussed in the appendixes: control of radiation damage in a graphitc reactor; annealing of graphite moderator structure in the BNL; annealing operation in BNL graphite reactor; effect of pile radiation on mechanical and other properties of graphite; neutron sensing instrumentation; instrumentation for sensing fuel failures; thermocouple pattern for enriched fuel loading; environmental hazard from a molten fuel element; retention of volatile flssion products on filters; retention of volatile fission products on water tube coolers; retention of volatile fission products in molten fuel plates; and release of the lowtemperature stored energy in the BEPO Pile. (W.L.H.)

1959-10-31T23:59:59.000Z

230

Evaluation of neutron absorbers for the melt-dilute treatment of aluminum-based spent fuel  

SciTech Connect

Aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors is being consolidated at the Savannah River Site (SRS) for ultimate disposal in the Mined Geologic Disposal System (MGDS). Most of the aluminum-based fuel material contains highly enriched uranium (HEU) (greater than 20 percent {sup 235}U), which poses a proliferation risk and challenges the preclusion of criticality events for disposal periods exceeding 10,000 years.

Vinson, D.W.

2000-03-31T23:59:59.000Z

231

A Review of Previous Research in Direct Energy Conversion Fission Reactors  

DOE Green Energy (OSTI)

From the earliest days of power reactor development, direct energy conversion was an obvious choice to produce high efficiency electric power generation. Directly capturing the energy of the fission fragments produced during nuclear fission avoids the intermediate conversion to thermal energy and the efficiency limitations of classical thermodynamics. Efficiencies of more than 80% are possible, independent of operational temperature. Direct energy conversion fission reactors would possess a number of unique characteristics that would make them very attractive for commercial power generation. These reactors would be modular in design with integral power conversion and operate at low pressures and temperatures. They would operate at high efficiency and produce power well suited for long distance transmission. They would feature large safety margins and passively safe design. Ideally suited to production by advanced manufacturing techniques, direct energy conversion fission reactors could be produced more economically than conventional reactor designs. The history of direct energy conversion can be considered as dating back to 1913 when Moseleyl demonstrated that charged particle emission could be used to buildup a voltage. Soon after the successful operation of a nuclear reactor, E.P. Wigner suggested the use of fission fragments for direct energy conversion. Over a decade after Wigner's suggestion, the first theoretical treatment of the conversion of fission fragment kinetic energy into electrical potential appeared in the literature. Over the ten years that followed, a number of researchers investigated various aspects of fission fragment direct energy conversion. Experiments were performed that validated the basic physics of the concept, but a variety of technical challenges limited the efficiencies that were achieved. Most research in direct energy conversion ceased in the US by the late 1960s. Sporadic interest in the concept appears in the literature until this day, but there have been no recent significant programs to develop the technology.

DUONG,HENRY; POLANSKY,GARY F.; SANDERS,THOMAS L.; SIEGEL,MALCOLM D.

1999-09-22T23:59:59.000Z

232

REACTOR  

DOE Patents (OSTI)

A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

Roman, W.G.

1961-06-27T23:59:59.000Z

233

U.S. Department of Energy Instrumentation and Controls Technology Research for Advanced Small Modular Reactors  

Science Conference Proceedings (OSTI)

Instrumentation, controls, and human-machine interfaces (ICHMI) are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The U.S. Department of Energy (DOE) has recognized that ICHMI research, development, and demonstration (RD&D) is needed to resolve the technical challenges that may compromise the effective and efficient utilization of modern ICHMI technology and consequently inhibit realization of the benefits offered by expanded utilization of nuclear power. Consequently, key DOE programs have substantial ICHMI RD&D elements to their respective research portfolio. This article describes current ICHMI research to support the development of advanced small modular reactors.

Wood, Richard Thomas [ORNL

2012-01-01T23:59:59.000Z

234

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

235

Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research  

SciTech Connect

The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called User’s Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. User’s week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

John Jackson; Todd Allen; Frances Marshall; Jim Cole

2013-03-01T23:59:59.000Z

236

A neutronic feasibility study for LEU conversion of the Budapest research reactor.  

SciTech Connect

A neutronic feasibility study for conversion of the Budapest Research Reactor (BRR) from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with the KFKI Atomic Energy Research Institute in Hungary. Comparisons were made of the reactor performance with the current HEU (36%) fuel and with a proposed LEU (19.75%) fuel. Cycle lengths, thermal neutron fluxes, and rod worths were calculated in equilibrium-type cores for each type of fuel. Relative to the HEU fuel, the LEU fuel has up to a 50% longer fuel cycle length, but a 7-10% smaller thermal neutron flux in the experiment locations. The rod worths are smaller with the LEU fuel, but are still large enough to easily satisfy the BRR shutdown margin criteria. Irradiation testing of four VVR-M2 LEU fuel assemblies that are nearly the same as the proposed BRR LEU fuel assemblies is currently in progress at the Petersburg Nuclear Physics Institute.

Pond, R. B.

1998-10-16T23:59:59.000Z

237

DEVELOPMENT OF A FAST-RELEASE ELECTRO-MAGNET FOR POOL-TYPE RESEARCH REACTORS  

SciTech Connect

Experimental data and observations of the design parameters and physical configurations which lead to faster-release cylindrical flat-faced electromagnets are given. Detailed drawings and operating characteristics are presented for an electromagnet suited to the requirements of a pool-type research reactor using either gravity drop or additional accelerating force. This electromagnet embodies the experimentally developed techniques for attaining fast release. (auth)

Michelson, C.

1958-02-14T23:59:59.000Z

238

Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report  

SciTech Connect

Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

1982-03-01T23:59:59.000Z

239

A simple setup for neutron tomography at the Portuguese Nuclear Research Reactor  

E-Print Network (OSTI)

A simple setup for neutron radiography and tomography was recently installed at the Portuguese Research Reactor. The objective of this work was to determine the operational characteristics of the installed setup, namely the irradiation time to obtain the best dynamic range for individual images and the spatial resolution. The performance of the equipment was demonstrated by imaging a fragment of a 17th century decorative tile.

M. A. Stanojev Pereira; J. G. Marques; R. Pugliesi

2012-05-15T23:59:59.000Z

240

Facts and issues of direct disposal of spent fuel; Revision 1  

Science Conference Proceedings (OSTI)

This report reviews those facts and issues that affect the direct disposal of spent reactor fuels. It is intended as a resource document for those impacted by the current Department of Energy (DOE) guidance that calls for the cessation of fuel reprocessing. It is not intended as a study of the specific impacts (schedules and costs) to the Savannah River Site (SRS) alone. Commercial fuels, other low enriched fuels, highly enriched defense-production, research, and naval reactor fuels are included in this survey, except as prevented by rules on classification.

Parks, P.B.

1993-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" 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

Reactor physics teaching and research in the Swiss nuclear engineering master  

Science Conference Proceedings (OSTI)

Since 2008, a Master of Science program in Nuclear Engineering (NE) has been running in Switzerland, thanks to the combined efforts of the country's key players in nuclear teaching and research, viz. the Swiss Federal Inst.s of Technology at Lausanne (EPFL) and at Zurich (ETHZ), the Paul Scherrer Inst. (PSI) at Villigen and the Swiss Nuclear Utilities (Swissnuclear). The present paper, while outlining the academic program as a whole, lays emphasis on the reactor physics teaching and research training accorded to the students in the framework of the developed curriculum. (authors)

Chawla, R. [Swiss Federal Inst. of Technology EPFL, CH-1015 Lausanne (Switzerland); Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland)

2012-07-01T23:59:59.000Z

242

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

243

IGORR-1: Proceedings of the first meeting of the international group on research reactors  

SciTech Connect

Many organizations, in several countries, are planning or implementing new or upgraded research reactor projects, but there has been no organized forum devoted entirely to discussion and exchange of information in this field. Over the past year or so, informal discussions resulted in widespread agreement that such a forum would serve a useful purpose. Accordingly, a proposal to form a group was submitted to the leading organizations known to be involved in projects to build or upgrade reactor facilities. Essentially all agreed to join in the formation of the International Group on Research Reactors (IGORR) and nominated a senior staff member to serve on its international organizing committee. The first IGORR meeting took place on February 28--March 2, 1990. It was very successful and well attended; some 52 scientists and engineers from 25 organizations in 10 countries participated in 2-1/2 days of open and informative presentations and discussions. Two workshop sessions offered opportunities for more detailed interaction among participants and resulted in identification of common R D needs, sources of data, and planned new facilities. Individual papers have been cataloged separately.

West, C.D. (comp.)

1990-05-01T23:59:59.000Z

244

HTGR spent fuel storage study  

SciTech Connect

This report documents a study of alternate methods of storing high-temperature gas-cooled reactor (HTGR) spent fuel. General requirements and design considerations are defined for a storage facility integral to a fuel recycle plant. Requirements for stand-alone storage are briefly considered. Three alternate water-cooled storage conceptual designs (plug well, portable well, and monolith) are considered and compared to a previous air-cooled design. A concept using portable storage wells in racks appears to be the most favorable, subject to seismic analysis and economic evaluation verification.

Burgoyne, R.M.; Holder, N.D.

1979-04-01T23:59:59.000Z

245

Evaluation of the thermal-hydraulic operating limits of the HEU-LEU transition cores for the MIT Research Reactor  

E-Print Network (OSTI)

The MIT Research Reactor (MITR) is in the process of conducting a design study to convert from High Enrichment Uranium (HEU) fuel to Low Enrichment Uranium (LEU) fuel. The currently selected LEU fuel design contains 18 ...

Wang, Yunzhi (Yunzhi Diana)

2009-01-01T23:59:59.000Z

246

Thermal hydraulics analysis of the MIT research reactor in support of a low enrichment uranium (LEU) core conversion  

E-Print Network (OSTI)

The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density monolithic UMo fuel. The design of an optimum LEU core for the ...

Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

2008-01-01T23:59:59.000Z

247

NUMERICAL SIMULATION FOR MECHANICAL BEHAVIOR OF U10MO MONOLITHIC MINIPLATES FOR RESEARCH AND TEST REACTORS  

Science Conference Proceedings (OSTI)

This article presents assessment of the mechanical behavior of U-10wt% Mo (U10Mo) alloy based monolithic fuel plates subject to irradiation. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Identification of the stress/strain characteristics is important for understanding the in-reactor performance of these plate-type fuels. For this work, three distinct cases were considered: (1) fabrication induced residual stresses (2) thermal cycling of fabricated plates; and finally (3) transient mechanical behavior under actual operating conditions. Because the temperatures approach the melting temperature of the cladding during the fabrication and thermal cycling, high temperature material properties were incorporated to improve the accuracy. Once residual stress fields due to fabrication process were identified, solution was used as initial state for the subsequent simulations. For thermal cycling simulation, elasto-plastic material model with thermal creep was constructed and residual stresses caused by the fabrication process were included. For in-service simulation, coupled fluid-thermal-structural interaction was considered. First, temperature field on the plates was calculated and this field was used to compute the thermal stresses. For time dependent mechanical behavior, thermal creep of cladding, volumetric swelling and fission induced creep of the fuel foil were considered. The analysis showed that the stresses evolve very rapidly in the reactor. While swelling of the foil increases the stress of the foil, irradiation induced creep causes stress relaxation.

Hakan Ozaltun & Herman Shen

2011-11-01T23:59:59.000Z

248

ORNL/TM-2012/380 Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research  

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

2/380 2/380 Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program September 2012 Prepared by Cyrus Smith Randy Nanstad Robert Odette Dwight Clayton Katie Matlack Pradeep Ramuhalli Glenn Light DOCUMENT AVAILABILITY Reports produced after January 1, 1996, are generally available free via the U.S. Department of Energy (DOE) Information Bridge. Web site http://www.osti.gov/bridge Reports produced before January 1, 1996, may be purchased by members of the public from the following source. National Technical Information Service 5285 Port Royal Road Springfield, VA 22161 Telephone 703-605-6000 (1-800-553-6847) TDD 703-487-4639 Fax 703-605-6900

249

Neutronic safety parameters and transient analyses for Poland's MARIA research reactor.  

SciTech Connect

Reactor kinetic parameters, reactivity feedback coefficients, and control rod reactivity worths have been calculated for the MARIA Research Reactor (Swierk, Poland) for M6-type fuel assemblies with {sup 235}U enrichments of 80% and 19.7%. Kinetic parameters were evaluated for family-dependent effective delayed neutron fractions, decay constants, and prompt neutron lifetimes and neutron generation times. Reactivity feedback coefficients were determined for fuel Doppler coefficients, coolant (H{sub 2}O) void and temperature coefficients, and for in-core and ex-core beryllium temperature coefficients. Total and differential control rod worths and safety rod worths were calculated for each fuel type. These parameters were used to calculate generic transients for fast and slow reactivity insertions with both HEU and LEU fuels. The analyses show that the HEU and LEU cores have very similar responses to these transients.

Bretscher, M. M.; Hanan, N. A.; Matos, J. E.; Andrzejewski, K.; Kulikowska, T.

1999-09-27T23:59:59.000Z

250

RELAP5 Application to Accident Analysis of the NIST Research Reactor  

Science Conference Proceedings (OSTI)

Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accident and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.

Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.; Diamond, D.

2012-03-18T23:59:59.000Z

251

Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have concluded, however, that with adequate engineered cooling of the vessel, the A508/533 steels are also acceptable.

J. K. Wright; R. N. Wright

2008-04-01T23:59:59.000Z

252

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

253

Initiation of a phase-I trial of neutron capture therapy at the MIT research reactor  

Science Conference Proceedings (OSTI)

The Massachusetts Institute of Technology (MIT), the New England Medical Center (NEMC), and Boston University Medical Center (BUMC) initiated a phase-1 trial of boron neutron capture therapy (BNCT) on September 6, 1994, at the 5-MW(thermal) MIT research reactor (MITR). A novel form of experimental cancer therapy, BNCT is being developed for certain types of highly malignant brain tumors such as glioblastoma and melanoma. The results of the phase-1 trials on patients with tumors in the legs or feet are described.

Harling, O.K.; Bernard, J.A.; Yam, Chun-Shan [Massachussets Institute of Technology, Cambridge, MA (United States)] [and others

1995-12-31T23:59:59.000Z

254

Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices  

SciTech Connect

Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

1982-03-01T23:59:59.000Z

255

Evaluation of differential shim rod worth measurements in the Oak Ridge Research Reactor  

Science Conference Proceedings (OSTI)

Reasonable agreement between calculated and measured differential shim rod worths in the Oak Ridge Research Reactor (ORR) has been achieved by taking into account the combined effects of negative reactivity contributions from changing fuel-moderator temperatures and of delayed photoneutrons. A method has been developed for extracting the asymptotic period from the shape of the initial portion of the measured time-dependent neutron flux profile following a positive reactivity insertion. In this region of the curve temperature-related reactivity feedback effects are negligibly small. Results obtained by applying this technique to differential shim rod worth measurements made in a wide variety of ORR cores are presented.

Bretscher, M.M.

1987-01-01T23:59:59.000Z

256

Preliminary study on direct recycling of spent PWR fuel in PWR system  

Science Conference Proceedings (OSTI)

Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jalan Ganesa 10 Bandung 40132 (Indonesia)

2012-06-06T23:59:59.000Z

257

Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

J. K. Wright; R. N. Wright

2010-07-01T23:59:59.000Z

258

Studies and research concerning BNFP. Spent fuel disassembly and canning program at Barnwell Nuclear Fuel Plant (BNFP) preliminary process assessment studies  

SciTech Connect

Studies being performed to assess nuclear fuel disassembly and encapsulation to enhance spent fuel storage have not revealed any conditions which unfavorably impact the feasibility of the concept. The studies are aimed at evaluating various issues warranting resolution preliminary to licensing a facility for this spent fuel management concept. The areas assessed are potential accidents and their results; maximum temperature level of canned fuel rods; radiation exposure to personnel during operation and fuel storage; criticality risks during storage and as a result of abnormal incidents; physical security requirements and material accountability measures; the impact of safeguards on economics; and a license schedule projection.

Tepp, H.G.

1980-10-01T23:59:59.000Z

259

Program on Technology Innovation: Volume Reduction Methods and Waste Form Changes for High-Activity Spent Resin  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) has initiated a series of studies to mitigate the impact of limited disposal-site access on continued light water reactor operations. A previous EPRI report, Program on Technology Innovation: Volume Reduction Methods and Waste Form Changes for High-Activity Spent Resin: A Feasibility Study (1025303), established that cation and anion resin beads could be separated for the purpose of rendering the anion resin as Class A resin waste, and ...

2013-11-14T23:59:59.000Z

260

Investigation of the condition of spent-fuel pool components  

Science Conference Proceedings (OSTI)

It is currently projected that spent nuclear fuel, which is discharged from the reactor and then stored in water pools, may remain in those pools for several decades. Other studies have addressed the expected integrity of the spent fuel during extended water storage; this study assesses the integrity of metallic spent fuel pool components. Results from metallurgical examinations of specimens taken from stainless steel and aluminum components exposed in spent fuel pools are presented. Licensee Event Reports (LERs) relating to problems with spent fuel components were assessed and are summarized to define the types of operational problems that have occurred. The major conclusions of this study are: aluminum and stainless steel spent fuel pool components have a good history of performance in both deionized and borated water pools. Although some operational problems involving pool components have occurred, these problems have had minimal impacts.

Kustas, F.M.; Bates, S.O.; Opitz, B.E.; Johnson, A.B. Jr.; Perez, J.M. Jr.; Farnsworth, R.K.

1981-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" 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

Nondestructive verification and assay systems for spent fuels  

SciTech Connect

This is an interim report of a study concerning the potential application of nondestructive measurements on irradiated light-water-reactor (LWR) fuels at spent-fuel storage facilities. It describes nondestructive measurement techniques and instruments that can provide useful data for more effective in-plant nuclear materials management, better safeguards and criticality safety, and more efficient storage of spent LWR fuel. In particular, several nondestructive measurement devices are already available so that utilities can implement new fuel-management and storage technologies for better use of existing spent-fuel storage capacity. The design of an engineered prototype in-plant spent-fuel measurement system is approx. 80% complete. This system would support improved spent-fuel storage and also efficient fissile recovery if spent-fuel reprocessing becomes a reality.

Cobb, D.D.; Phillips, J.R.; Bosler, G.E.; Eccleston, G.W.; Halbig, J.K.; Hatcher, C.R.; Hsue, S.T.

1982-04-01T23:59:59.000Z

262

Bioremediation of Diesel Contaminated Soil Using Spent Mushroom Compost.  

E-Print Network (OSTI)

??Composting has been shown to be an effective bioremediation technique for the treatment of hydrocarbon-contaminated soil. In this research, spent mushroom compost (SMC), a sustainable,… (more)

Eramo, Alessia

2009-01-01T23:59:59.000Z

263

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

264

Thermal-hydraulic calculations for the conversion to LEU of a research reactor core  

SciTech Connect

The thermal-hydraulic analysis performed for the needs of the conversion of the open pool 5MW Greek Research Reactor (GRR-1) to a pure Low Enrichment (LEU) configuration is presented. The methodology was based on a complete set of neutronic calculations performed for the new core configuration, in compliance with pre-defined Operation Limiting Conditions. The hottest channel analysis approach was adopted, and peaking factors were used to account for fabrication or measuring uncertainties. Calculations were carried out using the numerical codes NATCON, PLTEMP and PARET provided by Argonne National Laboratory (ANL). Two main different classes of conditions were considered, namely i) steady state normal operating conditions and ii) transient cases related to accidental events including reactivity feedback effects. For steady state operating conditions the behaviour of the new configuration was examined both for forced and natural convection cooling modes. Transient calculations considered several initiating events including reactivity insertion accidents (slow or fast reactivity insertion) and total or partial loss-of-flow accidents, i.e. in accordance to guidelines provided by the IAEA for research Reactors. (author)

Grigoriadis, D. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece); Department of Mechanical and Manufacturing Engineering, University of Cyprus, P.O. Box 20537, Nicosia 1678 (Cyprus); Varvayanni, M.; Catsaros, N.; Stakakis, E. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece)

2008-07-15T23:59:59.000Z

265

An analysis of radionuclide behavior in water pools during accidents at the Annular Core Research Reactor  

SciTech Connect

Physical and chemical phenomena that will affect the behavior of radionuclides released from fuel in the Annular Core Research Reactor during a hypothetical, core disruptive accident are described. The phenomena include boiling of water on heated clad, metal-water reactions, vapor nucleation to form aerosol particles, coagulation of aerosol particles, aerosol deposition within bubbles rising through the shield pool, vapor dissolution in the shield pool, and revaporization of radionuclides from the shield pool. A model of these phenomena is developed and applied to predict the release of radionuclides to the confinement building of the Annular Core Research Reactor. It is found that the shield pool provides overall decontamination factors for particulate of about 2.8 {times} 10{sup 5} and decontamination factors for noble gases of about 2.5--3.7. These results are found to be sensitive to the predicted clad temperature and bubble behavior in the shield pool. Slow revalorization of krypton, xenon and iodine from the shield pool is shown to create a prolonged, low-intensity source term of radioactive material to the confinement atmosphere.

Powers, D.A.

1992-08-01T23:59:59.000Z

266

A neutronic feasibility study for LEU conversion of the Brookhaven Medical Research Reactor (BMRR).  

SciTech Connect

A neutronic feasibility study for converting the Brookhaven Medical Research Reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with Brookhaven National Laboratory. Two possible LEU cores were identified that would provide nearly the same neutron flux and spectrum as the present HEU core at irradiation facilities that are used for Boron Neutron Capture Therapy and for animal research. One core has 17 and the other has 18 LEU MTR-type fuel assemblies with uranium densities of 2.5g U/cm{sup 3} or less in the fuel meat. This LEU fuel is fully-qualified for routine use. Thermal hydraulics and safety analyses need to be performed to complete the feasibility study.

Hanan, N. A.

1998-01-14T23:59:59.000Z

267

Supercritical Water Reactor (SCWR) - Survey of Materials Research and Development Needs to Assess Viability  

SciTech Connect

Supercritical water-cooled reactors (SCWRs) are among the most promising advanced nuclear systems because of their high thermal efficiency [i.e., about 45% vs. 33% of current light water reactors (LWRs)] and considerable plant simplification. SCWRs achieve this with superior thermodynamic conditions (i.e., high operating pressure and temperature), and by reducing the containment volume and eliminating the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers. The reference SCWR design in the U.S. is a direct cycle, thermal spectrum, light-water-cooled and moderated reactor with an operating pressure of 25 MPa and inlet/outlet coolant temperature of 280/500 °C. The inlet flow splits, partly to a down-comer and partly to a plenum at the top of the reactor pressure vessel to flow downward through the core in special water rods to the inlet plenum. This strategy is employed to provide good moderation at the top of the core, where the coolant density is only about 15-20% that of liquid water. The SCWR uses a power conversion cycle similar to that used in supercritical fossil-fired plants: high- intermediate- and low-pressure turbines are employed with one moisture-separator re-heater and up to eight feedwater heaters. The reference power is 3575 MWt, the net electric power is 1600 MWe and the thermal efficiency is 44.8%. The fuel is low-enriched uranium oxide fuel and the plant is designed primarily for base load operation. The purpose of this report is to survey existing materials for fossil, fission and fusion applications and identify the materials research and development needed to establish the SCWR viabilitya with regard to possible materials of construction. The two most significant materials related factors in going from the current LWR designs to the SCWR are the increase in outlet coolant temperature from 300 to 500 °C and the possible compatibility issues associated with the supercritical water environment. • Reactor pressure vessel • Pumps and piping

Philip E. MacDonald

2003-09-01T23:59:59.000Z

268

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

269

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

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

Why is the High Flux Beam Reactor Being Decommissioned? Why is the High Flux Beam Reactor Being Decommissioned? HFBR The High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is being decommissioned because the Department of Energy (DOE) decided in 1999 that it would be permanently closed. The reactor was shut down in 1997 after tritium from a leak in the spent-fuel pool was found in the groundwater. The HFBR, which had operated from 1965 to 1996, was used solely for scientific research, providing neutrons for materials science, chemistry, biology, and physics experiments. The reactor was shut down for routine maintenance in November of 1996. In January 1997, tritium, a radioactive form of hydrogen and a by-product of reactor operations, was found in groundwater monitoring wells immediately south of the HFBR. The tritium

270

N-Reactor Department Research and Development budget for FY 1966 and revision of budget for FY 1965  

SciTech Connect

The N-Reactor Department Research and Development Program for FY 1965, 1966, and later years is structured to achieve the following general goals. (1) Assurance of a high level of nuclear safety; (2) Assurance of achieving full plant life; (3) Reduction in operating costs for a given production rate; (4) Increase in production rate without proportionate increase in operating costs; (5) Savings in capital outlays necessary to achieve stated reductions in operating cost or increases in production; (6) Production of new products of value; (7) Savings in capital outlays or operating costs to achieve a given level of plant safety. The program is divided into three general categories; Reactor, Metallurgy, and Co-Product. The Reactor category is further divided into physics studies, thermal hydraulics studies, zircaloy process tube development, control, instrument and system analyses, chemistry, engineering research and development, gas, atmosphere studies, graphite studies, and nuclear safety research.

1964-03-25T23:59:59.000Z

271

A neutronic feasibility study for LEU conversion of the WWR-SM research reactor in Uzbekistan.  

SciTech Connect

The WWR-SM research reactor in Uzbekistan has operated at 10 MW since 1979, using Russian-supplied IRT-3M fuel assemblies containing 90% enriched uranium. Burnup tests of three full-sized IRT-3M FA with 36% enrichment were successfully completed to a burn up of about {approximately}50% in 1987-1989. In August 1998, four IRT-3M FA with 36% enriched uranium were loaded into the core to initiate conversion of the entire core to 36% enriched fuel. This paper presents the results of equilibrium fuel cycle comparisons of the reactor using HEU (90%) and HEU (36%) IRT-3M fuel and compares results with the performance of IRT-4M FA containing LEU (19.75%). The results show that an LEU (19.75%) density of 3.8 g/cm{sup 3} is required to match the cycle length of the HEU (90%) core and an LEU density 3.9 g/cm{sup 3} is needed to match the cycle length of the HEU (36%) core.

Rakhmanov, A.

1998-10-19T23:59:59.000Z

272

A neutronic feasibility study for LEU conversion of the IR-8 research reactor.  

SciTech Connect

Equilibrium fuel cycle comparisons for the IR-8 research reactor were made for HEU(90%), HEU(36%), and LEU (19.75%) fuel assembly (FA) designs using three dimensional multi-group diffusion theory models benchmarked to detailed Monte Carlo models of the reactor. Comparisons were made of changes in reactivity, cycle length, average {sup 235}U discharge burnup, thermal neutron flux, and control rod worths for the 90% and 36% enriched IRT-3M fuel assembly and the 19.75% enriched IRT-4M fuel assembly with the same fuel management strategy. The results of these comparisons showed that a uranium density of 3.5 g/cm{sup 3} in the fuel meat would be required in the LEU IRT-4M fuel assembly to match the cycle length of the HEU(90%) IRT-3M FA and an LEU density of 3.7 g/cm{sup 3} is needed to match the cycle length of the HEU(36%) IRT-3M FA.

Deen, J. R.

1998-10-22T23:59:59.000Z

273

Progressive Application Decommissioning Models for U.S. Power and Research Reactors  

SciTech Connect

This paper presents progressive engineering techniques and experiences in decommissioning projects performed by Bums and Roe Enterprises within the last fifteen years. Specifically, engineering decommissioning technical methods and lessons learned are discussed related to the Trojan Large Component Removal Project, San Onofre Nuclear Generating Station (SONGS) Decommissioning Project and the Brookhaven Graphite Research Reactor (BGRR) Decommissioning Project Study. The 25 years since the 1979 TMI accident and the events following 9/11 have driven the nuclear industry away from excessive, closed/elitist conservative methods towards more pragmatic results-oriented and open processes. This includes the essential recognition that codes, standards and regulatory procedures must be efficient, effective and fit for purpose. Financial and open-interactive stakeholder pressures also force adherence to aggressive risk reduction posture in the area of a safety, security and operations. The engineering methods and techniques applied to each project presented unique technical solutions. The decommissioning design for each project had to adopt existing design rules applicable to construction of new nuclear power plants and systems. It was found that the existing ASME, NRC, and DOE codes and regulations for deconstruction were, at best, limited or extremely conservative in their applicability to decommissioning. This paper also suggests some practical modification to design code rules in application for decommissioning and deconstruction. The representative decommissioning projects, Trojan, SONGS and Brookhaven, are discussed separately and the uniqueness of each project, in terms of engineering processes and individual deconstruction steps, is discussed. Trojan Decommissioning. The project included removal of entire NSSS system. The engineering complexity was mainly related to the 1200 MW Reactor. The approach, process of removal, engineering method related to protect the worker against excessive radiation exposure, transportation, and satisfying applicable rules and regulations, were the major problems to overcome. The project's successful completed earned a patent award. SONGS Decommissioning. The reactor's spherical containment and weakened integrity was the scope of this decommissioning effort. The aspects of structure stability and method of deconstruction is the major part of the presentation. The economical process of deconstruction, aspects of structural stability, worker safety, and the protection of the surrounding environment from contamination is highlighted in this section. BGRR Decommissioning Study. BREI was commissioned by Brookhaven National Laboratory (BNL) to evaluate and analyze the stability, and progressive decommissioning, and removal of BGRR components. This analysis took the form of several detailed decommissioning studies that range from disassembly and removal of the unit's graphite pile to the complete environmental restoration of the reactor site. While most of the facility's decommissioning effort is conventional, the graphite pile and its biological shield present the greatest challenge. The studies develop a unique method of removing high-activity waste trapped in the graphite joints. (authors)

Studnicka, Z.; Lacy, N.H.; Nicholas, R.G.; Campagna, M.; Morgan, R.D. [Bums and Roe Enterprises, Inc., 800 Kinderkamack Road, Oradell, NJ 07649 (United States); Sawruk, W. [ABS Consulting, Inc., 5 Birdsong Court, Shillington, PA 19607 (United States)

2006-07-01T23:59:59.000Z

274

Fuel development activities of the US RERTR Program. [Reduced Enrichment Research and Test Reactor  

SciTech Connect

Progress in the development and irradiation testing of high-density fuels for use with low-enriched uranium in research and test reactors is reported. Swelling and blister-threshold temperature data obtained from the examination of miniature fuel plates containing UAl/sub x/, U/sub 3/O/sub 8/, U/sub 3/Si/sub 2/, or U/sub 3/Si dispersed in an aluminum matrix are presented. Combined with the results of metallurgical examinations, these data show that these four fuel types will perform adequately to full burnup of the /sup 235/U contained in the low-enriched fuel. The exothermic reaction of the uranium-silicide fuels with aluminum has been found to occur at about the same temperature as the melting of the aluminum matrix and cladding and to be essentially quenched by the melting endotherm. A new series of miniature fuel plate irradiations is also discussed.

Snelgrove, J.L.; Domagala, R.F.; Wiencek, T.C.; Copeland, G.L.

1983-01-01T23:59:59.000Z

275

Transient analyses and thermal-hydraulic safety margins for the Greek Research Reactor (GRRI)  

Science Conference Proceedings (OSTI)

Various core configurations for the Greek research reactor (GRR1) have been considered in assessing the safety issues of adding a beryllium reflector to the existing water reflected HEU core and the transition from HEU to an all LEU core. The assessment has included both steady-state and transient analyses of safety margins and limits. A small all fresh Be reflected HEU core with a rather large nuclear peaking factor can still be operated safely, and thus adding a Be reflector to the larger depleted HEU core should not pose a problem. The transition mixed core with 50% LEU elements has larger void and Doppler coefficients than the HEU reference core and gives a lower peak clad temperature under transient conditions. The transition cores should give ever increasing margins to plate melting and fission product release as LEU elements are added to the core.

Woodruff, W.L.; Deen, J.R. [Argonne National Lab., IL (United States); Papastergiou, C. [National Centre for Scientific Research Demokritos, Athens (Greece)

1995-02-01T23:59:59.000Z

276

Reactor pressure vessel integrity research at the Oak Ridge National Laboratory  

Science Conference Proceedings (OSTI)

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the integrity inherent in the RPV. For this reason, the U.S. Nuclear Regulatory Commission has established the related research programs at ORNL described herein to provide for the development and confirmation of the methods used for: (1) establishing the irradiation exposure conditions within the RPV in the Embrittlement Data Base and Dosimetry Evaluation Program, (2) assessing the effects of irradiation on the RPV materials in the Heavy-Section Steel Irradiation Program, and (3) developing overall structural and fracture analyses of RPVs in the Heavy-Section Steel Technology Program.

Corwin, W.R.; Pennell, W.E.; Pace, J.V.

1995-12-31T23:59:59.000Z

277

Neutronic safety and transient analyses for potential LEU conversion of the IR-8 research reactor.  

SciTech Connect

Kinetic parameters, isothermal reactivity feedback coefficients and three transients for the IR-8 research reactor cores loaded with either HEU(90%), HEU(36%), or LEU (19.75%) fuel assemblies (FA) were calculated using three dimensional diffusion theory flux solutions, RELAP5/MOD3.2 and PARET. The prompt neutron generation time and effective delayed neutron fractions were calculated for fresh and beginning-of-equilibrium-cycle cores. Isothermal reactivity feedback coefficients were calculated for changes in coolant density, coolant temperature and fuel temperature in fresh and equilibrium cores. These kinetic parameters and reactivity coefficients were used in transient analysis models to predict power histories, and peak fuel, clad and coolant temperatures. The transients modeled were a rapid and slow loss-of-flow, a slow reactivity insertion, and a fast reactivity insertion.

Deen, J. R.; Hanan, N. A.; Smith, R. S.; Matos, J. E.; Egorenkov, P. M.; Nasonov, V. A.

1999-09-27T23:59:59.000Z

278

Preliminary neutronics calculations for conversion of the Tehran research reactor core from HEU to LEU fuel  

SciTech Connect

The 5-MW highly enriched uranium (HEU)-fueled Tehran Research Reactor is considered for conversion to high-density, low-enriched uranium (LEU) fuel. A preliminary neutronics calculation is performed as part of the conversion goal. In this study, two cores are considered: the HEU reference core and a proposed LEU core similar to the reference core, and a proposed LEU core similar to the reference core, using standardized U[sub 3]Si[sub 2] plates with the option of different [sup 235]U loadings. Various possibilities are investigated for the conversion of HEU to LEU fuel elements with 20% enriched [sup 235]U loadings of 207 to 297 g [sup 235]U/element. For the same equilibrium cycle length, the fuels are compared for flux, power distribution, burnup, and reactivity.

Nejat, S.M.R. (McMaster Univ., Hamilton, Ontario (Canada). Dept. of Engineering Physics.)

1993-08-01T23:59:59.000Z

279

Characterization of wastes in and around early reactors at the Hanford Site: The use of historical research  

SciTech Connect

This paper will present the waste characterization knowledge that has been gained in the first, ``Large-Scale Remediation Study`` to be performed on the reactor areas (100 Areas) of the Hanford Site. Undertaken throughout the past year, this research project has identified thousands of pieces of buried hardware, as well as the volumes of liquid wastes in burial sites in the reactor areas. The author of this landmark study, Dr. Michele Gerber, will discuss historical research as a safe and cost-effective characterization tool.

Gerber, M.S.

1993-10-01T23:59:59.000Z

280

Spent Fuel Transportation Risk Assessment  

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

Fuel Transportation Risk Assessment Fuel Transportation Risk Assessment (SFTRA) Draft NUREG-2125 Overview for National Transportation Stakeholders Forum John Cook Division of Spent Fuel Storage and Transportation 1 SFTRA Overview Contents * Project and review teams * Purpose and goals * Basic methodology * Improvements relative to previous studies * Draft NUREG structure and format * Routine shipment analysis and results * Accident condition analysis and results * Findings and conclusions * Schedule 2 SFTRA Research and Review Teams * Sandia National Laboratory Research Team [$1.8M; 9/06-9/12] - Doug Ammerman - principal investigator - Carlos Lopez - thermal - Ruth Weiner - RADTRAN * NRC's SFTRA Technical Review Team - Gordon Bjorkman - structural

Note: This page contains sample records for the topic "research reactor spent" 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

Baseline descriptions for LWR spent fuel storage, handling, and transportation  

SciTech Connect

Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables.

Moyer, J.W.; Sonnier, C.S.

1978-04-01T23:59:59.000Z

282

C-1 coproduct development for N Reactor: Excerpts from N-Reactor Department research and development budget for FY-1967 and revision of budget for FY-1966  

SciTech Connect

This report discusses the production of weapon grade plutonium and tritium as an economically attractive mode of operation for N-Reactor. With the successful completion of the proposed research and development program coproduct production in N-Reactor could begin in January 1967. Approximately 5 kg of tritium could be produced at a reduction in unit total cost of up to $7 per gram equivalent when compared to other modes of operation of N-Reactor that yield only-weapon grade plutonium. Achieving this production level and unit cost requires a fuel and target design that will subject the materials to more severe temperature and exposure conditions than for the tube-in-tube design. Analyses of existing knowledge and experimental results obtained to date are very encouraging. However, there is a continuing need for additional experimental verification of the limiting conditions for both fuel and target to permit the specification of a production fuel and target design. The incremental cost of this program is estimated to be $1,055 million in FY-1965, $0.796 million in FY-1966 and $0.450 million in FY-1967 for a total of $2.3 million. The principal benefits to be derived from this expenditure of research and development funds are lower unit grade plutonium. When the reactor reaches equilibrium operation as a costs for both tritium and weapon dual-purpose reactor producing process steam for credit, it will be a highly competitive producer of plutonium and tritium in the AEC complex.

1965-04-19T23:59:59.000Z

283

Very High Temperature Reactor (VHTR) Survey of Materials Research and Development Needs to Support Early Deployment  

DOE Green Energy (OSTI)

The VHTR reference concept is a helium-cooled, graphite moderated, thermal neutron spectrum reactor with an outlet temperature of 1000 C or higher. It is expected that the VHTR will be purchased in the future as either an electricity producing plant with a direct cycle gas turbine or a hydrogen producing (or other process heat application) plant. The process heat version of the VHTR will require that an intermediate heat exchanger (IHX) and primary gas circulator be located in an adjoining power conversion vessel. A third VHTR mission - actinide burning - can be accomplished with either the hydrogen-production or gas turbine designs. The first ''demonstration'' VHTR will produce both electricity and hydrogen using the IHX to transfer the heat to either a hydrogen production plant or the gas turbine. The plant size, reactor thermal power, and core configuration will be designed to assure passive decay heat removal without fuel damage during accidents. The fuel cycle will be a once-through very high burnup low-enriched uranium fuel cycle. The purpose of this report is to identify the materials research and development needs for the VHTR. To do this, we focused on the plant design described in Section 2, which is similar to the GT-MHR plant design (850 C core outlet temperature). For system or component designs that present significant material challenges (or far greater expense) there may be some viable design alternatives or options that can reduce development needs or allow use of available (cheaper) materials. Nevertheless, we were not able to assess those alternatives in the time allotted for this report and, to move forward with this material research and development assessment, the authors of this report felt that it was necessary to use a GT-MHR type design as the baseline design.

Eric Shaber; G. Baccaglini; S. Ball; T. Burchell; B. Corwin; T. Fewell; M. Labar; P. MacDonald; P. Rittenhouse; Russ Vollam; F. Southworth

2003-01-01T23:59:59.000Z

284

Fuel cycle design and analysis of SABR: subrcritical advanced burner reactor.  

E-Print Network (OSTI)

??Various fuel cycles for a sodium-cooled, subcritical, fast reactor with a fusion neutron source for the transmutation of light water reactor spent fuel have been… (more)

Sommer, Christopher

2008-01-01T23:59:59.000Z

285

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

286

Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.  

SciTech Connect

This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics computer codes and methodologies applied to SSTAR. Another section of the report details the SSTAR safety design approach which is based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials and how the inherent safety features of the lead coolant, nitride fuel, fast neutron spectrum core, pool vessel configuration, natural circulation, and containment meet or exceed the requirements for each level of protection. The report also includes recent results of a systematic analysis by LANL of data on corrosion of candidate cladding and structural material alloys of interest to SSTAR by LBE and Pb coolants; the data were taken from a new database on corrosion by liquid metal coolants created at LANL. The analysis methodology that considers penetration of an oxidation front into the alloy and dissolution of the trailing edge of the oxide into the coolant enables the long-term corrosion rate to be extracted from shorter-term corrosion data thereby enabling an evaluation of alloy performance over long core lifetimes (e.g., 30 years) that has heretofore not been possible. A number of candidate alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR would operate were analyzed following testing in the DELTA loop at LANL including steels that were treated by laser peening at LLNL; laser peening is an approach that alters the oxide-metal bonds which could potentially improve corrosion resistance. LLNL is also carrying out Multi-Scale Modeling of the Fe-Cr system with the goal of assisting in the development of cladding and structural materials having greater resistance to irradiation.

Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

2008-06-23T23:59:59.000Z

287

Arrival condition of spent fuel after storage, handling, and transportation  

Science Conference Proceedings (OSTI)

This report presents the results of a study conducted to determine the probable arrival condition of spent light-water reactor (LWR) fuel after handling and interim storage in spent fuel storage pools and subsequent handling and accident-free transport operations under normal or slightly abnormal conditions. The objective of this study was to provide information on the expected condition of spent LWR fuel upon arrival at interim storage or fuel reprocessing facilities or at disposal facilities if the fuel is declared a waste. Results of a literature survey and data evaluation effort are discussed. Preliminary threshold limits for storing, handling, and transporting unconsolidated spent LWR fuel are presented. The difficulty in trying to anticipate the amount of corrosion products (crud) that may be on spent fuel in future shipments is also discussed, and potential areas for future work are listed. 95 references, 3 figures, 17 tables.

Bailey, W.J.; Pankaskie, P.J.; Langstaff, D.C.; Gilbert, E.R.; Rising, K.H.; Schreiber, R.E.

1982-11-01T23:59:59.000Z

288

Impact of PWR spent fuel variations on TRU-fueled VHTRS  

E-Print Network (OSTI)

Several alternative strategies are being considered as spent nuclear fuel (SNF) management options. Transuranic nuclides (TRU) are responsible for the SNF long-term radiotoxicity beyond the first 500 years. One of the most viable approaches suggests creating new transmutation fuels containing TRUs for use in thermal and fast nuclear reactors. Irradiation of TRUs results in their transmutation and ultimate incineration by fission. The objective of this thesis is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled Very High Temperature Reactor (VHTR) systems. This effort was focused on the prismatic core configuration. The 3D core models were created for use in calculations with the SCALE 5.1 code system. As part of the research effort, basic nuclear characteristics of TRUs were taken into consideration. The potential variations of PWR spent fuel compositions were modeled with the International Atomic Energy Agency (IAEA) Nuclear Fuel Cycle Simulation System, VISTA. The VHTR configurations with varying TRU compositions were analyzed assuming a single-batch core operation. Their performance was compared to the VHTR cases with low enriched uranium (LEU). The analysis shows that TRUs can be effectively utilized in the VHTR systems. The TRU-fueled VHTRs exhibit favorable performance characteristics.

Alajo, Ayodeji Babatunde

2007-12-01T23:59:59.000Z

289

Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one-on-one interviews and traditional and online focus groups with scientists. The latter include SNS, HFIR, and APS users as well as scientists at ORNL, some of whom had not yet used HFIR and/or SNS. These approaches informed development of the second phase, a quantitative online survey. The survey consisted of 16 questions and 7 demographic categorizations, 9 open-ended queries, and 153 pre-coded variables and took an average time of 18 minutes to complete. The survey was sent to 589 SNS/HFIR users, 1,819 NSLS users, and 2,587 APS users. A total of 899 individuals provided responses for this study: 240 from NSLS; 136 from SNS/HFIR; and 523 from APS. The overall response rate was 18%.

Bryant, Rebecca [Bryant Research, LLC; Kszos, Lynn A [ORNL

2011-03-01T23:59:59.000Z

290

Optimization of irradiation conditions for {sup 177}Lu production at the LVR-15 research reactor  

Science Conference Proceedings (OSTI)

The use of lutetium in medicine has been increasing over the last few years. The {sup 177}Lu radionuclide is commercially available for research and test purposes as a diagnostic and radiotherapy agent in the treatment of several malignant tumours. The yield of {sup 177}Lu from the {sup 176}Lu(n,{gamma}){sup 177}Lu nuclear reaction depends significantly on the thermal neutron fluence rate. The capture cross-sections of both reaction {sup 176}Lu(n,{gamma}){sup 177}Lu and reaction {sup 177}Lu(n,{gamma}){sup 178}Lu are very high. Therefore a burn-up of target and product nuclides should be taken into account when calculating {sup 177}Lu activity. The maximum irradiation time, when the activity of the {sup 177}Lu radionuclide begins to decline, was found for different fluence rates. Two vertical irradiation channels at the LVR-15 nuclear research reactor were compared in order to choose the channel with better irradiation conditions, such as a higher thermal neutron fluence rate in the irradiation volume. In this experiment, lutetium was irradiated in a titanium capsule. The influence of the Ti capsule on the neutron spectrum was monitored using activation detectors. The choice of detectors was based on requirements for irradiation time and accurate determination of thermal neutrons. The following activation detectors were selected for measurement of the neutron spectrum: Ti, Fe, Ni, Co, Ag and W. (authors)

Lahodova, Z.; Viererbl, L.; Klupak, V. [Research Centre Rez Ltd., Husinec-130, Rez 250 67 (Czech Republic); Srank, J. [Nuclear Physics Inst. of the Academy of Sciences, Husinec-130, Rez 250 67 (Czech Republic)

2012-07-01T23:59:59.000Z

291

Final Site Specific Decommissioning Inspection Report #2 for the University of Washington Research and Test Reactor, Seattle, Washington  

SciTech Connect

During the period of August through November 2006, ORISE performed a comprehensive IV at the University of Washington Research and Test Reactor Facility. The objective of the ORISE IV was to validate the licensee’s final status survey processes and data, and to assure the requirements of the DP and FSSP were met.

S.J. Roberts

2007-03-20T23:59:59.000Z

292

Comparison of shell-and-tube with frame-and-plate-type heat exchangers for the MIT research reactor upgrade  

SciTech Connect

This paper discusses a comparison of shell-and-tube with frame-and-plate-type heat exchangers for the proposed upgrade of the Massachusetts Institute of Technology research reactor (MITR). The comparison is based on the following considerations: thermal-hydraulic performance, maintenance, personnel dose rate, and pricing.

Ser, Yorick, Hu, Lin-Wen [Massachusetts Institute of Technology, Cambridge, MA (United States)

1997-12-01T23:59:59.000Z

293

Analysis of in-core experiment activities for the MIT Research Reactor using the ORIGEN computer code  

E-Print Network (OSTI)

The objective of this study is to devise a method for utilizing the ORIGEN-S computer code to calculate the activation products generated in in-core experimental assemblies at the MIT Research Reactor (MITR-II). ORIGEN-S ...

Helvenston, Edward M. (Edward March)

2006-01-01T23:59:59.000Z

294

Spent fuel test project, Climax granitic stock, Nevada Test Site  

SciTech Connect

The Spent Fuel Test-Climax (SFT-C) is a test of dry geologic storage of spent nuclear reactor fuel. The SFT-C is located at a depth of 420 m in the Climax granitic stock at the Nevada Test Site. Eleven canisters of spent commercial PWR fuel assemblies are to be stored for 3 to 5 years. Additional heat is supplied by electrical heaters, and more than 800 channels of technical information are being recorded. The measurements include rock temperature, rock displacement and stress, joint motion, and monitoring of the ventilation air volume, temperature, and dewpoint.

Ramspott, L.D.

1980-10-24T23:59:59.000Z

295

Further Evaluation of the Neutron Resonance Transmission Analysis (NRTA) Technique for Assaying Plutonium in Spent Fuel  

Science Conference Proceedings (OSTI)

This is an end-of-year report (Fiscal Year (FY) 2011) for the second year of effort on a project funded by the National Nuclear Security Administration's Office of Nuclear Safeguards (NA-241). The goal of this project is to investigate the feasibility of using Neutron Resonance Transmission Analysis (NRTA) to assay plutonium in commercial light-water-reactor spent fuel. This project is part of a larger research effort within the Next-Generation Safeguards Initiative (NGSI) to evaluate methods for assaying plutonium in spent fuel, the Plutonium Assay Challenge. The second-year goals for this project included: (1) assessing the neutron source strength needed for the NRTA technique, (2) estimating count times, (3) assessing the effect of temperature on the transmitted signal, (4) estimating plutonium content in a spent fuel assembly, (5) providing a preliminary assessment of the neutron detectors, and (6) documenting this work in an end of the year report (this report). Research teams at Los Alamos National Laboratory (LANL), Lawrence Berkeley National Laboratory (LBNL), Pacific Northwest National Laboratory (PNNL), and at several universities are also working to investigate plutonium assay methods for spent-fuel safeguards. While the NRTA technique is well proven in the scientific literature for assaying individual spent fuel pins, it is a newcomer to the current NGSI efforts studying Pu assay method techniques having just started in March 2010; several analytical techniques have been under investigation within this program for two to three years or more. This report summarizes work performed over a nine month period from January-September 2011 and is to be considered a follow-on or add-on report to our previous published summary report from December 2010 (INL/EXT-10-20620).

J. W. Sterbentz; D. L. Chichester

2011-09-01T23:59:59.000Z

296

DISPERSIONS OF URANIUM CARBIDES IN ALUMINUM PLATE-TYPE RESEARCH REACTOR FUEL ELEMENTS  

DOE Green Energy (OSTI)

The technical feasibility of employing uranium carbide aluminun dispersions in aluminum-base research reactor fuel elements was investigated This study was motivated by the need to obtain higher uranium loadings in these fuel elements. Although toe MTR-type unit, containing a 13 18 wt% U-Al alloy is a proven reactor component, fabrication problems of considerable magnitude arise when attempts are made to increase the uranium investment in the alloy to more than 25 wt.%. Au approach to these fabrication difficulties is to select a compound with significantly higher density tban UAl/sub 4/ or UAl/sub 3/ compounds of the alloy system which when dispersed in aluminum powder, will reduce the volume occupied by the brittle, fissile phase. The uranium carbides, with densities ranging from 11.68 to 13.63 g/cm/sup 3/), appear to be suited for this application and were selected for development as a fuel material for aluminum-base dispersions. Studies were conducted at 580 to 620 deg C to determine the chemical compatibility of carbides with aluminum in sub-size cold- pressed comparts as well as in full-size fabricated fuel plates. Procedures were also developed to prepare uranium carbides, homogernously disperse the compounds in aluminum, roll clad the dispersions to form composite plates, and braze the plates into fuel assemblies. Corrosion tests of the fuel material were conducted in 20 and 60 deg C water to determine the integrity of the fuel material in the event of sin inadventent cladding failure. In addition, specimens were prepared to evaluate penformance under extensive irradiation Prior to studying the uranium carbide-aluminum system, methods for preparing the carbides were investigated. Are melting uranium and carnon was satisfactory for obtaining small quantities of various carbides. Later, reaction of graphite with UO/sub 2/ was successfully employed in the preparation of large quantities of UC/sub 2/, Studies of the chemical compatibility of cold-pressed compacts containing 50 wt% uranium carbide dispersed in aluminum revealed a marked trend toward stebifity as the carbon content of the uranium carbide increased from 446 to 9.20% C. Severe volume increases occurred in monocarbide dispersions with attendant formation of large quantities of the uranium-allumnim inter-metallic compounds. Dicarbide dispersions, on the other band, exhibited negligible reaction with aluminum after extended periods at 580 and 620 deg C. However, it was demonstrated that hydrogen can promote a reaction in UC/sub 2/-Al compacts. The hydrogen appears to reduce the UC/sub 2/ to UC which can subsequently react with aluminum producing the previously noted deleterious effects. A growth study at 605 deg C of composite fuel plates containing 59 wt.% UC/sub 2/ revealed insignificant changes within processing periods envisioned for fuel element processing. However, plate elongations as high as 2.5% were observed after 100 hr at this temperature. Severe blistering which occurred on fuel plates fabricated in the initial stages of the investigation was attributed to gaseous hydrocarbons, and the condition was ellminated by vacuum degasification of cold-pressed compacts. With the exception of the degasification requirement, procedures for manufacturing UC- bearing fuel elements were identical to those specified for the Geneva Conference Reactor fuel elements. Dispersions of uranium dicarbide corroded catastrophically in 20 and 60 deg C water, thus limiting the application of this material However, spocimens were prepared and insented in the MTR to evaluate the irradiation behavior of this fuel because of its potential application in onganic- cooled reactors. (auth)

Thurber, W.C.; Beaver, R.J.

1959-11-19T23:59:59.000Z

297

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

298

License amendment for neutron capture therapy at the MIT research reactor  

SciTech Connect

This paper reports the issuance by the U.S. Nuclear Regulatory Commission (NRC) of a license amendment to the Massachusetts Institute of Technology (MIT) for the use of the MIT Research Reactor's (MITR-II) medical therapy facility beam for the treatment of humans using neutron capture therapy (NCT). This amendment is one of 11 required approvals. The others are those of internal MIT committees, review panels of the Tufts-New England Medical Center (NEMC), which is directing the program jointly with MIT, that of the U.S. Food and Drug Administration, and an NRC amendment to the NEMC hospital license. This amendment is the first of its type to be issued by NRC, and as such it establishes a precedent for the conduct of human therapy using neutron beams. Neutron capture therapy is a bimodal method for treating cancer that entails the administration of a tumor-seeking boronated drug followed by the irradiation of the target organ with neutrons. The latter cause boron nuclei to fission and thereby release densely ionizing helium and lithium nuclei, which destroy cancerous cells while leaving adjacent healthy cells undamaged. Neutron capture therapy is applicable to glioblastoma multiforme (brain tumors) and metastasized melanoma (skin cancer). Both Brookhaven National Laboratory and MIT conducted trials of NCT more than 30 yr ago. These were unsuccessful because the available boron drugs did not concentrate sufficiently in tumor and because the thermal neutron beams that were used did not enable neutrons to travel deep enough into the brain.

Bernard, J.A. (Massachusetts Institute of Technology, Cambride, MA (United States))

1993-01-01T23:59:59.000Z

299

A technical review of non-destructive assay research for the characterization of spent nuclear fuel assemblies being conducted under the US DOE NGSI  

SciTech Connect

There is a growing belief that expansion of nuclear energy generation will be needed in the coming decades as part of a mixed supply chain to meet global energy demand. At stake is the health of the economic engine that delivers human prosperity. As a consequence renewed interest is being paid to the safe management of spent nuclear fuel (SNF) and the plutonium it contains. In addition to being an economically valuable resource because it can be used to construct explosive devices, Pu must be placed on an inventory and handled securely. A multiinstitutional team of diverse specialists has been assembled under a project funded by the US Department of Energy (DOE) Next Generation Safeguards Initiative (NGSI) to address ways to nondestructively quantify the plutonium content of spent nuclear fuel assemblies, and to also detect the potential diversion of pins from those assemblies. Studies are underway using mostly Monte Carlo tools to assess the feasibility, individual and collective performance capability of some fourteen nondestructive assay methods. Some of the methods are familiar but are being applied in a new way against a challenging target which is being represented with a higher degree of realism in simulation space than has been done before, while other methods are novel. In this work we provide a brief review of the techniques being studied and highlight the main achievements to date. We also draw attention to the deficiencies identified in for example modeling capability and available basic nuclear data. We conclude that this is an exciting time to be working in the NDA field and that much work, both fundamental and applied, remains ahead if we are to advance the state of the practice to meet the challenges posed to domestic and international safeguards by the expansion of nuclear energy together with the emergence of alternative fuel cycles.

Croft, Stephen [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory

2010-12-06T23:59:59.000Z

300

Light Water Reactor Sustainability Research and Development Program Plan -- Fiscal Year 2009–2013  

SciTech Connect

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. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R&D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R&D programs. The purpose of the LWRS R&D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R&D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its international engagement and leadership on nuclear safety and security issues.

Idaho National Laboratory

2009-12-01T23:59:59.000Z

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301

Light Water Reactor Sustainability Research and Development Program Plan -- Fiscal Year 2009–201/span>3  

Science Conference Proceedings (OSTI)

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. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R&D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R&D programs. The purpose of the LWRS R&D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R&D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its international engagement and leadership on nuclear safety and security issues.

Idaho National Laboratory

2009-12-01T23:59:59.000Z

302

Electrorefining {open_quotes}N{close_quotes} reactor fuel  

SciTech Connect

Principles of purifying of uranium metal by electrorefining are reviewed. Metal reactor fuel after irradiation is a form of impure uranium. Dissolution and deposition electrorefining processes were developed for spent metal fuel under the Integral Fast Reactor Program. Application of these processes to the conditioning of spent N-reactor fuel slugs is examined.

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

1995-02-01T23:59:59.000Z

304

Inspection methods for physical protection. Task II. Review of research reactor licensees' physical security practices  

SciTech Connect

Security systems and security procedures for the AFRRI reactor, the University of Maryland TRIGA reactor, and the University of Virginia CAVALIER and UVAR reactors are described.

1980-07-03T23:59:59.000Z

305

Cost Impact of Using ISG-8 Rev. 3 for PWR Spent Fuel Pool Criticality Analysis  

Science Conference Proceedings (OSTI)

Nuclear Regulatory Commission (NRC) guidance for applying burnup credit in criticality analyses for spent fuel storage and transportation requirements recently changed with the release of Interim Staff Guidance (ISG) 8 Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks. If ISG-8 Rev. 3 were imposed upon pressurized water reactor (PWR) spent fuel pool (SFP) criticality analyses, the burnup requirements for loading would ...

2012-11-21T23:59:59.000Z

306

Gas Reactor Technology R&D  

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

U.S. Department of Energy to Invest U.S. Department of Energy to Invest up to $7.3 Million for "Deep-Burn" Gas-Reactor Technology R&D Artist's rendering of Nuclear Plant An artist's rendering of the Next Generation Nuclear Plant concept. The U.S. Department of Energy today announced a Funding Opportunity Announcement (FOA) valued at $7.3 million for universities, commercial entities, National Laboratories with expertise in the concept of nuclear fuel "Deep-Burn" in which plutonium and higher transuranics recycled from spent nuclear fuel are destroyed. The funding opportunity seeks to establish the technological foundations that will support the role of the very-high-temperature, gas-cooled reactor (VHTR) in the nuclear fuel cycle -- which is one of the prototype reactors being researched/developed under

307

Epithermal beam development at the BMRR (Brookhaven Medical Research Reactor): Dosimetric evaluation  

SciTech Connect

The utilization of an epithermal neutron beam for neutron capture therapy (NCT) is desirable because of the increased tissue penetration relative to a thermal neutron beam. Over the past few years, modifications have been and continue to be made at the Brookhaven Medical Research Reactor (BMRR) by changing its filter components to produce an optimal epithermal beam. An optimal epithermal beam should contain a low fast neutron contamination and no thermal neutrons in the incident beam. Recently a new moderator for the epithermal beam has been installed at the epithermal port of the BMRR and has accomplished this task. This new moderator is a combination of alumina (Al{sub 2}O{sub 3}) bricks and aluminum (Al) plates. A 0.51 mm thick cadmium (Cd) sheet has reduced the thermal neutron intensity drastically. Furthermore, an 11.5 cm thick bismuth (Bi) plate installed at the port surface has reduced the gamma dose component to negligible levels. Foil activation techniques have been employed by using bare gold and cadmium-covered gold foil to determine thermal as well as epithermal neutron fluence. Fast neutron fluence has been determined by indium foil counting. Fast neutron and gamma dose in soft tissue, free in air, is being determined by the paired ionization chamber technique, using tissue equivalent (TE) and graphite chambers. Thermoluminescent dosimeters (TLD-700) have also been used to determine the gamma dose independently. This paper describes the methods involved in the measurements of the above mentioned parameters. Formulations have been developed and the various corrections involved have been detailed. 12 refs.

Saraf, S.K.; Fairchild, R.G.; Kalef-Ezra, J.; Laster, B.H.; Fiarman, S.; Ramsey, E. (Brookhaven National Lab., Upton, NY (USA); Ioannina Univ. (Greece); Brookhaven National Lab., Upton, NY (USA); State Univ. of New York, Stony Brook, NY (USA). Health Science Center)

1989-08-24T23:59:59.000Z

308

Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion  

SciTech Connect

Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

2012-09-30T23:59:59.000Z

309

Novel, Magnetically Fluidized-Bed Reactor Development for the Looping Process: Coal to Hydrogen Production Research and Development  

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

Novel, Magnetically Fluidized-Bed Novel, Magnetically Fluidized-Bed Reactor Development for the Looping Process: Coal to Hydrogen Production Research and Development Background The U.S. Department of Energy (DOE) National Energy Technology Laboratory (NETL) is committed to improving methods for co-producing power and chemicals, fuels, and hydrogen (H2). Gasification is a process by which fuels such as coal can be used to produce synthesis gas (syngas), a mixture of H2, carbon monoxide (CO), and carbon

310

Conceptual Design of Molten Salt Loop Experiment for MIT Research Reactor  

E-Print Network (OSTI)

Molten salt is a promising coolant candidate for Advanced High Temperature Reactor (AHTR) Gen-IV designs. The low neutron absorption, high thermal capacity, chemical inertness, and high boiling point at low pressure of ...

Bean, Malcolm K.

2011-08-01T23:59:59.000Z

311

Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior  

SciTech Connect

This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

Monteleone, S. [comp.

1995-04-01T23:59:59.000Z

312

Development of refined MCNPX-PARET multi-channel model for transient analysis in research reactors  

SciTech Connect

Reactivity insertion transients are often analyzed (RELAP, PARET) using a two-channel model, representing the hot assembly with specified power distribution and an average assembly representing the remainder of the core. For the analysis of protected by the reactor safety system transients and zero reactivity feedback coefficients this approximation proves to give adequate results. However, a more refined multi-channel model representing the various assemblies, coupled through the reactivity feedback effects to the whole reactor core is needed for the analysis of unprotected transients with excluded over power and period trips. In the present paper a detailed multi-channel PARET model has been developed which describes the reactor core in different clusters representing typical BR2 fuel assemblies. The distribution of power and reactivity feedback in each cluster of the reactor core is obtained from a best-estimate MCNPX calculation using the whole core geometry model of the BR2 reactor. The sensitivity of the reactor response to power, temperature and energy distributions is studied for protected and unprotected reactivity insertion transients, with zero and non-zero reactivity feedback coefficients. The detailed multi-channel model is compared vs. simplified fewer-channel models. The sensitivities of transient characteristics derived from the different models are tested on a few reactivity insertion transients with reactivity feedback from coolant temperature and density change. (authors)

Kalcheva, S.; Koonen, E. [SCK-CEN, BR2 Reactor Dept., Boeretang 200, 2400 Mol (Belgium); Olson, A. P. [RERTR Program, Nuclear Engineering Div., Argonne National Laboratory, Cass Avenue, Argonne, IL 60439 (United States)

2012-07-01T23:59:59.000Z

313

Thermal reclamation of spent blasting abrasive  

Science Conference Proceedings (OSTI)

Abrasive blasting media is used to remove anticorrosive and antifoulant coatings from the hulls and tanks of US Navy ships. The total production of paint-contaminated spent abrasives from the eight US. Navy shipyards ranges from 75,000 to 100,000 tons per year. Most of this spent abrasive is disposed in landfills. Organic paint binders and heavy metals are present in the spent abrasives in concentrations sufficient to classify them as hazardous wastes in some states. In an effort to avoid the rising costs an long-term environmental liability associated with landfilling this waste, the US Navy has investigated various methods of reclaiming spent abrasives for reuse in hull- and tank-blasting operations. This paper discusses the results of a research and development project conducted under the Navy's Hazardous Waste Minimization Program to test a fluidized-bed sloped-grid (FBSG) reclaimer to determine if it could be used to recycle spent abrasive. Thirty tons of abrasive were processed and a product meeting military specifications for new abrasives was reclaimed. Blasting performance was also comparable to new abrasives. 3 refs., 1 fig., 2 tabs.

Bryan, B.G. (Institute of Gas Technology, Chicago, IL (USA)); Thomas, W.; Adema, C. (David Taylor Research Center, Annapolis, MD (USA))

1990-01-01T23:59:59.000Z

314

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...

315

J. Plasma Fusion Res. SERIES, Vol. 10 (2013) Flibe-Tritium Research for Fission or Fusion Reactors at Kyushu University  

E-Print Network (OSTI)

There is increasing interest in using ionic molten-salt Flibe not only as self-cooled tritium(T)-breeding material in a fusion reactor blanket but also as fuel solvent of molten-salt fission reactors. Application of Flibe to T-breeding fluid for a stellarator-type fusion reactor operated at a high magnetic field brings large simplification of its blanket structure, allowing continuous operation under high-beta plasma conditions. Using mixed Flibe-ThF 4+UF 4 fuel in molten salt fission reactors permits stable long-term operation without fuel exchange. When Flibe or Flinak is irradiated by neutrons, however, acid and corrosive TF is generated, and some T permeates through structural walls. In order to solve these problems, chemical conditions of Flibe are changed using the redox-control reaction, Be+2TF=BeF 2+T 2. In addition, permeation of hydrogen isotopes is lowered by enhancing T recovery rates. Part of Flibe-tritium researches are performed at Idaho National Laboratory (INL) under the Japan-US collaboration work of JUPITER-II. Our own contributions to the topics are shortly introduced in this paper.

Satoshi Fukada

2012-01-01T23:59:59.000Z

316

Loss of spent fuel pool cooling PRA: Model and results  

Science Conference Proceedings (OSTI)

This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 {times} 10{sup {minus}5} and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 {times} 10{sup {minus}3}. Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible.

Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

1996-09-01T23:59:59.000Z

317

Revision to the Record of Decision for the Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel (DOE/EIS-218) (7/19/00)  

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

767 767 Federal Register / Vol. 65, No. 139 / Wednesday, July 19, 2000 / Notices The office is located in the Pentagon which is guarded. RETENTION AND DISPOSAL: Records are kept until the person is deceased or the person seeks removal of information, whichever is sooner. SYSTEM MANAGER(S) AND ADDRESS: Chief of Naval Operations (N09BC), 2000 Navy Pentagon, Washington, DC 20350-2000. NOTIFICATION PROCEDURE: Individuals seeking to determine whether information about themselves is contained in this system should address written inquiries to the Chief of Naval Operations (N09BC), 2000 Navy Pentagon, Washington, DC 20350-2000. RECORD ACCESS PROCEDURES: Individuals seeking access to information about themselves contained in this system should address written inquiries to the Chief of Naval

318

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

319

Expanding and optimizing fuel management and data analysis capabilities of MCODE-FM in support of MIT research reactor (MITR-II) LEU conversion  

E-Print Network (OSTI)

Studies are underway in support of the MIT research reactor (MITR-II) conversion from high enriched Uranium (HEU) to low enriched Uranium (LEU), as required by recent non-proliferation policy. With the same core configuration ...

Horelik, Nicholas E. (Nicholas Edward)

2012-01-01T23:59:59.000Z

320

Development of a core design optimization tool and analysis in support of the planned LEU conversion of the MIT Research Reactor (MITR-II)  

E-Print Network (OSTI)

The MIT Research Reactor (MITR-II) is currently undergoing analysis for the planned conversion from high enriched uranium (HEU) to low enriched uranium (LEU), as part of a global effort to minimize the availability of ...

Connaway, Heather M. (Heather Moira)

2012-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" 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

PRELIMINARY HAZARDS SUMMARY REPORT ON THE BROOKHAVEN HIGH FLUX BEAM RESEARCH REACTOR  

SciTech Connect

The High Flux Beam Reactor, HFBR, is cooled, moderated, and reflected by heavy water and designed to produce 40 Mw with a total epithermal flux of ~1.6 X 10/sup 15/cm/sup -2/ sec/sup -1/ and a flector thermal maximum flux of 7 X 10/sup 14/ cm/sup -2/ sec/sup -1/, using a core formed by ETR plate-type fuel elements in a close-packed array. The hazards summary is given in terms of site description, reactor design, building design, plant operation, disposal of radioactive wastes and effluents, and safety analysis. (B.O.G.)

Hendrie, J.M.; Kouts, H.J.C.

1961-05-01T23:59:59.000Z

322

EXPERIMENT OPERATIONS PLAN FOR A LOSS-OF-COOLANT ACCIDENT SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR  

SciTech Connect

Pressurized water reactor loss-of-coolant accident phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship between the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. This document contains both experiment proposal and assembly proposal information. The intent of this document is to supply information required by the Chalk River Nuclear Laboratories (CRNL), and to identify the planned procedures and data that will be used both to establish readiness to proceed from one test phase to the next and to operate the experiment. Operating control settings and limits are provided for both experimenter systems and CRNL systems. A hazards review summarizes safety issues that have been addressed during the development of the experiment plan.

Russcher, G. E.; Cannon, L. W.; Goodman, R. L.; Hesson, G. M.; King, L. L.; McDuffie, P. N.; Marshall, R. K.; Nealley, C.; Pilger, J. P.; Mohr, C. L.

1981-04-01T23:59:59.000Z

323

Decommissioning Plan of the Musashi Reactor and Its Progress  

SciTech Connect

The Musashi Reactor is a TRIGA-II, tank-type research reactor, as shown in Table 1. The reactor had been operated at maximum thermal power level of 100 kW since first critical, January 30, 1963. Reactor operation was shut down due to small leakage of water from the reactor tank on December 21,1989. After shutdown, investigation of the causes, making plan of repair and discussions on restart or decommissioning had been done. Finally, decision of decommissioning was made in May, 2003. The initial plan of the decommissioning was submitted to the competent authority in January, 2004. Now, the reactor is under decommissioning. The plan of decommissioning and its progress are described. In conclusion: considering the status of undertaking plan of the waste disposal facility for the low level radioactive waste from research reactors, the phased decommissioning was selected for the Musashi Reactor. First phase of the decommissioning activities including the actions of permanent shutdown and delivering the spent nuclear fuels to US DOE was completed.

Tanzawa, Tomio [Atomic Energy Research Laboratory, Musashi Institute of Technology, Ozenji 971, Asao-ku, Kawasaki, 215-0013 (Japan)

2008-01-15T23:59:59.000Z

324

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

325

MIT Modular Pebble Bed Reactor (MPBR) A Summary of Research Activities and Accomplishments  

E-Print Network (OSTI)

operation #12;MIT MPBR Specifications Thermal Power 250 MW - 120 Mwe Target Thermal Efficiency 45 % Core for a pebble bed reactor power plant system with high efficiency and minimum capital cost ­ Net efficiency > 45;Plant With Space Frames #12;#12;For 1150 MW Electric Power Station Turbine Hall Boundary Admin Training

326

Spent Fuel Working Group Report. Volume 1  

SciTech Connect

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

O`Toole, T.

1993-11-01T23:59:59.000Z

327

Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.  

Science Conference Proceedings (OSTI)

This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying/tuning of lower-fidelity models, which now require costly experimental qualification for each different type of design

Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

2007-06-30T23:59:59.000Z

328

CASMO-2 spent-fuel-rack criticality analysis  

SciTech Connect

In recent years, utilities have needed to increase their spent-fuel storage capacity. Both Maine Yankee pressurized water reactor (PWR) and Vermont Yankee boiling water reactor (BWR) have increased their spent-fuel rack capacity by decreasing the canister center-to-center spacing while adding fixed poison. Licensing criticality analysis of such changes in spent-fuel rack design have been performed at Yankee Atomic Electric Co. (YAEC) using NITAWL-KENO-IV and the 123-group XSDRN library. However, KENO/Monte Carlo analysis has inherent drawbacks when applied to spent-fuel rack design and modification. These include statistical uncertainty and long computer time. In contrast, the transport theory code, CASMO-2, provides deterministic and fast criticality analysis. Also, since collapsed and transport-corrected cross sections are generated, PDQ can be used to analyze large array problems which are prohibitively expensive using KENO. In this work, the authors apply the CASMO-PDQ methodology to the Maine Yankee and Vermont Yankee high-density spent-fuel rack designs, and compare the final results against KENO.

Napolitano, D.G.; Heinrichs, D.P.; Gorski, J.P.

1986-01-01T23:59:59.000Z

329

Examples of the use of PSA in the design process and to support modifications at two research reactors  

Science Conference Proceedings (OSTI)

Many, if not most, of the world`s commercial nuclear power plants have been the subject of plant-specific probabilistic safety assessments (PSA). A growing number of other nuclear facilities as well as other types of industrial installations have been the focus of plant-specific PSAs. Such studies have provided valuable information concerning the nature of the risk of the individual facility and have been used to identify opportunities to manage that risk. This paper explores the risk management activities associated with two research reactors in the United States as a demonstration of the versatility of the use of PSA to support risk-related decision making.

Johnson, D.H.; Bley, D.C.; Lin, J.C. [PLG, Inc., Newport Beach, CA (United States); Ramsey, C.T.; Linn, M.A. [Oak Ridge National Lab., TN (United States)

1994-03-01T23:59:59.000Z

330

Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting; Volume 2, Severe accident research, Thermal hydraulics  

SciTech Connect

This three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21--23, 1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchersfrom CEC, China, Finland, France, Germany, Japan, Spain and Taiwan. Selected papers have been processed separately for inclusion in the Energy Science and Technology Database.

Weiss, A.J. [comp.] [Brookhaven National Lab., Upton, NY (United States)

1993-03-01T23:59:59.000Z

331

Fluid-Structure Interaction for Coolant Flow in Research-type Nuclear Reactors  

Science Conference Proceedings (OSTI)

The High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), is scheduled to undergo a conversion of the fuel used and this proposed change requires an extensive analysis of the flow through the reactor core. The core consists of 540 very thin and long fuel plates through which the coolant (water) flows at a very high rate. Therefore, the design and the flow conditions make the plates prone to dynamic and static deflections, which may result in flow blockage and structural failure which in turn may cause core damage. To investigate the coolant flow between fuel plates and associated structural deflections, the Fluid-Structure Interaction (FSI) module in COMSOL will be used. Flow induced flutter and static deflections will be examined. To verify the FSI module, a test case of a cylinder in crossflow, with vortex induced vibrations was performed and validated.

Curtis, Franklin G [ORNL; Ekici, Kivanc [ORNL; Freels, James D [ORNL

2011-01-01T23:59:59.000Z

332

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

333

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

334

Spent Fuel Transportation Package Performance Study - Experimental Design Challenges  

Science Conference Proceedings (OSTI)

Numerous studies of spent nuclear fuel transportation accident risks have been performed since the late seventies that considered shipping container design and performance. Based in part on these studies, NRC has concluded that the level of protection provided by spent nuclear fuel transportation package designs under accident conditions is adequate. [1] Furthermore, actual spent nuclear fuel transport experience showcase a safety record that is exceptional and unparalleled when compared to other hazardous materials transportation shipments. There has never been a known or suspected release of the radioactive contents from an NRC-certified spent nuclear fuel cask as a result of a transportation accident. In 1999 the United States Nuclear Regulatory Commission (NRC) initiated a study, the Package Performance Study, to demonstrate the performance of spent fuel and spent fuel packages during severe transportation accidents. NRC is not studying or testing its current regulations, a s the rigorous regulatory accident conditions specified in 10 CFR Part 71 are adequate to ensure safe packaging and use. As part of this study, NRC currently plans on using detailed modeling followed by experimental testing to increase public confidence in the safety of spent nuclear fuel shipments. One of the aspects of this confirmatory research study is the commitment to solicit and consider public comment during the scoping phase and experimental design planning phase of this research.

Snyder, A. M.; Murphy, A. J.; Sprung, J. L.; Ammerman, D. J.; Lopez, C.

2003-02-25T23:59:59.000Z

335

Nondestructive verification and assay systems for spent fuels. Technical appendixes  

SciTech Connect

Six technical appendixes are presented that provide important supporting technical information for the study of the application of nondestructive measurements to spent-fuel storage. Each appendix addresses a particular technical subject in a reasonably self-contained fashion. Appendix A is a comparison of spent-fuel data predicted by reactor operators with measured data from reprocessors. This comparison indicates a rather high level of uncertainty in previous burnup calculations. Appendix B describes a series of nondestructive measurements at the GE-Morris Operation Spent-Fuel Storage Facility. This series of experiments successfully demonstrated a technique for reproducible positioning of fuel assemblies for nondestructive measurement. The experimental results indicate the importance of measuring the axial and angular burnup profiles of irradiated fuel assemblies for quantitative determination of spent-fuel parameters. Appendix C is a reasonably comprehensive bibliography of reports and symposia papers on spent-fuel nondestructive measurements to April 1981. Appendix D is a compendium of spent-fuel calculations that includes isotope production and depletion calculations using the EPRI-CINDER code, calculations of neutron and gamma-ray source terms, and correlations of these sources with burnup and plutonium content. Appendix E describes the pulsed-neutron technique and its potential application to spent-fuel measurements. Although not yet developed, the technique holds the promise of providing separate measurements of the uranium and plutonium fissile isotopes. Appendix F describes the experimental program and facilities at Los Alamos for the development of spent-fuel nondestructive measurement systems. Measurements are reported showing that the active neutron method is sensitive to the replacement of a single fuel rod with a dummy rod in an unirradiated uranium fuel assembly.

Cobb, D.D.; Phillips, J.R.; Baker, M.P.

1982-04-01T23:59:59.000Z

336

Evaluation of transuranium isotopes inventory for Candu/ACR standard and SEU spent fuel and the possibility to transmute them  

Science Conference Proceedings (OSTI)

Available in abstract form only. Full text of publication follows: The main disadvantage of nuclear energy is the quantity of long lived radioactive waste produced in a NPP. Transmutation could be one of the solutions to reduce it. Waste transmutation will require a suitable deployment of techniques for spent fuel reprocessing. At present, reprocessing is done by aqueous methods that are very efficient for Pu separation (up to 99.9%). For transmutation applications, new partitioning processes must be developed for minor actinides separation from the high level waste. Although these processes are still very much at the research stage, industrial scale-up will result in the deployment of new, more specific separation techniques for transmutation applications. Partitioning and Transmutation (P and T) techniques could contribute to reduce the radioactive inventory and its associated radio-toxicity. Scientists are looking for ways to drastically reduce both the mass and the radio-toxicity of the nuclear waste to be stored in a deep geological repository, and to reduce the time needed to reach the radioactivity level of the raw material originally used to produce energy. The first stage in the transmutation process is the isotopes inventory formed in the spent fuel. In this paper is made an intercomparison evaluation using WIMS 5B.12 and ORIGEN computer codes. Using these two codes, there is evaluated the isotopes released by a fuel standard from a Candu reactor. Moreover, there is simulated an inventory released by a Candu-SEU reactor and an ACR reactor. (authors)

Ghizdeanu, Elena Nineta; Pavelescu, Alexandru [University Politehnica of Bucharest - Faculty of Power Engineering, 313 Splaiul Independentei, RO-060042, Bucharest 6 (Romania); Balaceanu, Victoria [Institute for Nuclear Research, Campului Str., 1, Mioveni P.O. Box 78, 0300 Pitesti (Romania)

2007-07-01T23:59:59.000Z

337

Theoretical analysis of the subcritical experiments performed in the IPEN/MB-01 research reactor facility  

SciTech Connect

The theoretical analysis of the subcritical experiments performed at the IPEN/MB-01 reactor employing the coupled NJOY/AMPX-II/TORT systems was successfully accomplished. All the analysis was performed employing ENDF/B-VII.0. The theoretical approach follows all the steps of the subcritical model of Gandini and Salvatores. The theory/experiment comparison reveals that the calculated subcritical reactivity is in a very good agreement to the experimental values. The subcritical index ({xi}) shows some discrepancies although in this particular case some work still have to be made to model in a better way the neutron source present in the experiments. (authors)

Lee, S. M.; Dos Santos, A. [Inst. de Pesquisas Energeticas e Nucleares, Cidade Universitaria, Av. Lineu Prestes, 2242, 05508-000 Sao Paulo - SP (Brazil)

2012-07-01T23:59:59.000Z

338

Technology, safety, and costs of decommissioning reference nuclear research and test reactors: sensitivity of decommissioning radiation exposure and costs to selected parameters  

Science Conference Proceedings (OSTI)

Additional analyses of decommissioning at the reference research and test (R and T) reactors and analyses of five recent reactor decommissionings are made that examine some parameters not covered in the initial study report (NUREG/CR-1756). The parameters examined for decommissioning are: (1) the effect on costs and radiation exposure of plant size and/or type; (2) the effects on costs of increasing disposal charges and of unavailability of waste disposal capacity at licensed waste disposal facilities; and (3) the costs of and the available alternatives for the disposal of nuclear R and T reactor fuel assemblies.

Konzek, G.J.

1983-07-01T23:59:59.000Z

339

Actinide minimization using pressurized water reactors  

E-Print Network (OSTI)

Transuranic actinides dominate the long-term radiotoxity in spent LWR fuel. In an open fuel cycle, they impose a long-term burden on geologic repositories. Transmuting these materials in reactor systems is one way to ease ...

Visosky, Mark Michael

2006-01-01T23:59:59.000Z

340

Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel  

SciTech Connect

The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed.

Deen, J.R.; Snelgrove, J.L. (Argonne National Lab., IL (United States)); Papastergiou, C. (National Center for Scientific Research, Athens (Greece))

1992-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" 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

Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel  

SciTech Connect

The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed.

Deen, J.R.; Snelgrove, J.L. [Argonne National Lab., IL (United States); Papastergiou, C. [National Center for Scientific Research, Athens (Greece)

1992-12-31T23:59:59.000Z

342

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel...  

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

of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research...

343

A New Class of Risk-Importance Measures to Support Reactor Aging Management and the Prioritization of Materials Degradation Research  

Science Conference Proceedings (OSTI)

As the US fleet of light water reactors ages, the risks of operation might be expected to increase. Although probabilistic risk assessment has proven a critical resource in risk-informed regulatory decision-making, limitations in current methods and models have constrained their prospective value in reactor aging management. These limitations stem principally from the use of static component failure rate models (which do not allow the impact of component aging on failure rates to be represented) and a very limited treatment of passive components (which would be expected to have an increasingly significant risk contribution in an aging system). Yet, a PRA captures a substantial knowledge base that could be of significant value in addressing plant aging. In this paper we will describe a methodology and a new class of risk importance measures that allow the use of an existing PRA model to support the management of plant aging, the prioritization of improvements to non-destructive examination and monitoring techniques, and the establishment of research emphases in materials science. This methodology makes use of data resources generated under the USNRC Proactive Management of Materials Degradation program which addresses the anticipated effects of numerous aging degradation mechanisms on a wide variety of component types.

Unwin, Stephen D.; Lowry, Peter P.; Toyooka, Michael Y.

2010-06-07T23:59:59.000Z

344

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

345

METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS  

DOE Patents (OSTI)

A control arrangement is offered for a boiling-water reactor. Boric acid is maintained in the water in the reactor and the amount in the reactor is controlled by continuously removing a portion of the water from the reactor, concentrating the boric acid by evaporating the water therefrom, returning a controlled amount of the acid to the reactor, and simultaneously controlling the water level by varying the rate of spent steam return to the reactor.

Reed, G.A.

1961-01-10T23:59:59.000Z

346

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

347

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

348

Design of a low enrichment, enhanced fast flux core for the Massachusetts Institute of Technology Research Reactor  

E-Print Network (OSTI)

Worldwide, there is limited test reactor capacity to perform the required irradiation experiments on advanced fast reactor materials and fuel designs. This is particularly true in the U.S., which no longer has an operating ...

Ellis, Tyler Shawn

2009-01-01T23:59:59.000Z

349

Feasibility of Reduced Boron Concentration Operation in Pressurized Water Reactor Plants  

Science Conference Proceedings (OSTI)

Technical Paper / Second Seminar on Accelerated Testing of Materials in Spent Nuclear Fuel and High-Level Waste Storage Systems / Fission Reactors

Aung Tharn Daing; Myung-Hyun Kim

350

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

351

Summary of the EPRI Early Event Analysis of the Fukushima Daiichi Spent Fuel Pools Following the March 11, 2011 Earthquake and Tsuna mi in Japan  

Science Conference Proceedings (OSTI)

Damage to the Fukushima Daiichi Unit 4 reactor building observed on March 15, 2011, initially generated confusion and concern throughout the nuclear industry. The reactor had been defueled approximately 100 days prior to the March 11 earthquake and tsunami; therefore, any explosion in Unit 4 could not be linked to a recently operating reactor within that unit. With the full core in the spent fuel pool, suspicions immediately turned to hydrogen generated by oxidation of overheating spent fuel cladding fol...

2012-05-31T23:59:59.000Z

352

Intermodal transportation of spent fuel  

SciTech Connect

Concepts for transportation of spent fuel in rail casks from nuclear power plant sites with no rail service are under consideration by the US Department of Energy in the Commercial Spent Fuel Management program at the Pacific Northwest Laboratory. This report identifies and evaluates three alternative systems for intermodal transfer of spent fuel: heavy-haul truck to rail, barge to rail, and barge to heavy-haul truck. This report concludes that, with some modifications and provisions for new equipment, existing rail and marine systems can provide a transportation base for the intermodal transfer of spent fuel to federal interim storage facilities. Some needed land transportation support and loading and unloading equipment does not currently exist. There are insufficient shipping casks available at this time, but the industrial capability to meet projected needs appears adequate.

Elder, H.K.

1983-09-01T23:59:59.000Z

353

The Argonaut Reactor - Reactors designed/built by Argonne National  

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

Achievements > Achievements > Argonne Reactors > Training 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

354

Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs  

SciTech Connect

This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

Monteleone, S. [Brookhaven National Lab., Upton, NY (United States)] [comp.

1994-04-01T23:59:59.000Z

355

US Department of Energy Storage of Spent Fuel and High Level Waste  

DOE Green Energy (OSTI)

ABSTRACT This paper provides an overview of the Department of Energy's (DOE) spent nuclear fuel (SNF) and high level waste (HLW) storage management. Like commercial reactor fuel, DOE's SNF and HLW were destined for the Yucca Mountain repository. In March 2010, the DOE filed a motion with the Nuclear Regulatory Commission (NRC) to withdraw the license application for the repository at Yucca Mountain. A new repository is now decades away. The default for the commercial and DOE research reactor fuel and HLW is on-site storage for the foreseeable future. Though the motion to withdraw the license application and delay opening of a repository signals extended storage, DOE's immediate plans for management of its SNF and HLW remain the same as before Yucca Mountain was designated as the repository, though it has expanded its research and development efforts to ensure safe extended storage. This paper outlines some of the proposed research that DOE is conducting and will use to enhance its storage systems and facilities.

Sandra M Birk

2010-10-01T23:59:59.000Z

356

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

357

Spent Fuel Pool Risk Assessment Integration Framework (Mark I and II BWRs) and Pilot Plant Application  

Science Conference Proceedings (OSTI)

This report documents the development and pilot application of a generic framework and methodology for conducting a probabilistic risk assessment (PRA) for spent fuel pools at BWR plants with Mark I or II containment designs. A key aspect of the study is the consideration of potential synergistic relationships between adverse conditions in the reactor and the spent fuel pool.BackgroundUsed nuclear fuel from the operation of nuclear power plants is typically ...

2013-05-01T23:59:59.000Z

358

Quarterly technical report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, April--June 1975  

SciTech Connect

The current water reactor safety activities of ANC are accomplished in four programs. The Semiscale Program consists of small-scale nonnuclear thermal- hydraulic experiments for the generation of experimental data that can be applied to analytical models describing loss-of-coolant accident (LOCA) phenomena in water-cooled nuclear power plants. Emphasis is placed on acquiring system effects data from integral tests that characterize the most significant thermal- hydraulic phenomena during the depressurization (blowdown) and emergency cooling phase of a LOCA. The LOFT Program provides test data to support: (a) assessment and improvement of the analytical methods utilized for predicting the behavior of pressurized water reactors (PWR) under LOCA conditions; (b) evaluation of the performance of PWR engineered safety features (ESF), particularly the emergency core cooling system (ECCS); and (c) assessment of the quantitative margins of safety inherent in the performance of these safety features. The Thermal Fuels Behavior Program is a program designed to provide information on the behavior of reactor fuels under normal, off-normal, and accident conditions. The experimental portion is concentrated on testing of single fuel rods and fuel rod clusters under power-cooling-mismatch (PCM), loss-of-coolant, and reactivity initiated accident conditions. The Reactor Behavior Program encompasses the analytical aspects of predicting the response of nuclear power reactors under normal, abnormal, and accident conditions. The status of each program is reported. (auth)

1975-11-01T23:59:59.000Z

359

AN EVALUATION OF THE URANIUM CONTAMINATION ON THE SURFACES OF ALCLAD URANIUM-ALUMINUM ALLOY RESEARCH REACTOR FUEL PLATES  

SciTech Connect

Reported radioactivity in the Low-Intensity Test Reactor (LITR) water coolant traceable to uranium contamination on the surfaces of the alclad uranium-- aluminum plate-tyne fuel element led to an investigation to determine the sources of uranium contamination on the fuel plate surfaces. Two possible contributors to surface contamination are external sources such as rolling-mill equipment, the most obvious, and diffusion of uranium from the uranium-aluminum alloy fuel into the aluminum cladding. This diffusion is likely because of the 600 deg C heat treatments used in the conventional fabrication process. Uranium determinations based on neutron activation analysis of machined layers from fuel plate surfaces showed that rolling-mill equipment, contaminated with highly enriched uranium, was responsible for transferring as much as 180 ppm U to plate surfaces. By careful practice where cleanliness is emphasized, surface contamination can be reduced to 0.6 ppm U/sup 235/. The residue remaining on the plate surface may be accounted for by diffusion of uranium from the fuel alloy into and through the cladding of the fuel plate. Data obtained from preliminary diffusion studies permitted a good estimate to be made of the diffusion coefficient of uranium into aluminum at 600 deg C: 2.5 x 10/sup -8/ cm//sec. To minimize diffusion while the plate-type aluminum-base research reactor fuel element is being processed, heat treatments at 600 deg C should be limited to 2.5 hr. The uranium contamination on the surfaces of the finished fuel plates should then be less than 0.6 ppm U / sup 235/ . This investigation also revealed that the solubility limit of uranium in aluminum at 600 deg C is approx 60 ppm. (auth)

Beaver, R.J.; Erwin, J.H.; Mateer, R.S.

1962-03-19T23:59:59.000Z

360

M-1 N-Reactor fuel development. Excerpts from N-Reactor Department Research and Development Budget for FY-1967 and Revision of Budget for FY-1966  

SciTech Connect

The use of metallic uranium fuel elements on a large scale in a pressurized water cooled graphite moderated power reactor environment in N-Reactor not only represents an extrapolation from previous Hanford fuel technology but it represents the only production-scale application of this concept in this country or abroad. Provision of a supporting technology is and will continue to be a major activity at Hanford as only a limited amount of directly applicable R&D will be available from other sites. The major benefit to be achieved through an extensive and continuing fuel program is attainment of the full capabilities of the N-Reactor at much reduced fuel cycle costs. The current N-Reactor fuel design represents the best engineering judgment of what is required for adequate performance. While the design is intentionally conservative, some features may not fully provide the level of performance required to sustain efficient reactor operation. Results of production-scale irradiation experience and special test irradiations will provide direction to a continuing program to correct any excesses or deficiencies in the initial fuel design. Reduction of unwarranted conservatism in the design will lower fuel fabrication costs, and correction of deficiencies will lower irradiation costs through increased time operated efficiency. Costs of this program are summarized for 1965, 66, and 67. The paper describes the scope of the program, its relationship to other programs, and technical progress in FY-1965 which included improved performance of the fuel elements even with the large number of rapid shutdowns of the reactor.

1965-04-19T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" 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

Spent-fuel composition: a comparison of predicted and measured data  

Science Conference Proceedings (OSTI)

The uncertainty in predictions of the nuclear materials content of spent light-water reactor fuel was investigated to obtain guidelines for nondestructive spent-fuel verification and assay. Values predicted by the reactor operator were compared with measured values from fuel reprocessors for six reactors (three PWR and three BWR). The study indicates that total uranium, total plutonium, fissile uranium, fissile plutonium, and total fissile content can be predicted with biases ranging from 1 to 6% and variabilities (1-sigma) ranging from 2 to 7%. The higher values generally are associated with BWRs. Based on the results of this study, nondestructive assay measurements that are accurate and precise to 5 to 10% (1sigma) or better should be useful for quantitative analyses of typical spent fuel.

Thomas, C.C. Jr.; Cobb, D.D.; Ostenak, C.A.

1981-03-01T23:59:59.000Z

362

Applications of Nd:YAG laser micromanufacturing in High Temperature Gas Reactor research  

SciTech Connect

Two innovative applications of Nd:YAG laser micromachining techniques are demonstrated in this publication. Research projects to determine the fission product transport mechanisms in TRISO coated particles necessitate heat treatment studies as well as the manufacturing of a unique sealed system for experimentation at very high temperatures. This article describes firstly the design and creation of an alumina jig designed to contain 500 {mu}m diameter ZrO2 spheres intended for annealing experiments at temperatures up to 1600 C. Functional requirements of this jig are the precision positioning of spheres for laser ablation, welding and post weld heat treatment in order to ensure process repeatability and accurate indexing of individual spheres. The design challenges and the performance of the holding device are reported. Secondly the manufacture of a sealing system using laser micromachining is reported. ZrO2 micro plugs isolate the openings of micro-machined cavities to produce a gas-tight seal fit for application in a high temperature environment. The technique is described along with a discussion of the problems experienced during the sealing process. Typical problems experienced were seating dimensions and the relative small size ({approx} 200 {mu}m) of these plugs that posed handling challenges. Manufacturing processes for both the tapered seating cavity and the plug are demonstrated. In conclusion, this article demonstrates the application of Nd-YAG micromachining in an innovative way to solve practical research problems.

I. J. van Rooyen; C. A. Smal; J. Steyn; H. Greyling

2012-08-01T23:59:59.000Z

363

Seismic Safety Margins Research Programs. Assessment of potential increases in risk due to degradation of steam generator and reactor coolant pump supports. [PWR  

Science Conference Proceedings (OSTI)

During the NRC licensing review for the North Anna Units 1 and 2 pressurized-water reactors (PWRs), questions were raised regarding the potential for low-fracture toughness of steam-generator and reactor-coolant-pump supports. Because other PWRs may face similar problems, this issue was incorporated into the NRC Program for Resolution of Generic Issues. The work described in this report was performed to provide the NRC with a quantitative evaluation of the value/impact implications of the various options of resolving the fracture-toughness question. This report presents an assessment of the probabilistic risk associated with nil-ductility failures of steam-generator and reactor-coolant-pump structural-support systems during seismic events, performed using the Seismic Safety Margins Research Program codes and data bases.

Bohn, M. P.; Wells, J. E.; Shieh, L. C.; Cover, L. E.; Streit, R. L.

1983-08-01T23:59:59.000Z

364

The Integral Fast Reactor: A practical approach to waste management  

SciTech Connect

This report discusses development of the method for pyroprocessing of spent fuel from the Integral Fast Reactor (or Advanced Liquid Metal Reactor). The technology demonstration phase, in which recycle will be demonstrated with irradiated fuel from the EBR-II reactor has been reached. Methods for recovering actinides from spent LWR fuel are at an earlier stage of development but appear to be technically feasible at this time, and a large-scale demonstration of this process has begun. The utilization of fully compatible processes for recycling valuable spent fuel materials promises to provide substantial economic incentives for future applications of the pyroprocessing technology.

Laidler, J.J.

1993-12-31T23:59:59.000Z

365

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT  

SciTech Connect

Progress on reactor programs and in general engineering research and development programs is summarized. Research and development are reported on water-cooled reactors including EBWR and Borax-V, sodium-cooled reactors including ZPR-III, IV, and IX, Juggernaut, and EBR-I and II. Other work included a review of fast reactor technology, and studies on nuclear superheat, thermal and fast reactor safety, and reactor physics. Effort was also devoted to reactor materials and fuels development, heat engineering, separation processes and advanced reactor concepts. (J.R.D.)

1961-04-01T23:59:59.000Z

366

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT, MAY 1962  

SciTech Connect

Research progress is reported on water-cooled reactors, liquid-metal- cooled reactors, general reactor technology, plutonium recycle, advanced systems research and development, and nuclear safety. (M.C.G.)

1962-06-15T23:59:59.000Z

367

Spent-fuel-storage alternatives  

Science Conference Proceedings (OSTI)

The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

Not Available

1980-01-01T23:59:59.000Z

368

The Optimized Integration of the Decontamination Plan and the Radwaste Management Plan into Decommissioning Plan to the VVR-S Research Reactor from Romania  

SciTech Connect

The paper presents the progress of the Decontamination Plan and Radioactive Waste Management Plan which accompanies the Decommissioning Plan for research reactor VVR-S located in Magurele, Ilfov, near Bucharest, Romania. The new variant of the Decommissioning Plan was elaborated taking into account the IAEA recommendation concerning radioactive waste management. A new feasibility study for VVR-S decommissioning was also elaborated. The preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as rehabilitation of the existing Radioactive Waste Treatment Plant and the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and reusing of cleaned reactor building is envisaged. An inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the IAEA assistance. Environmental concerns are a part of the radioactive waste management strategy. In conclusion: The current version 8 of the Draft Decommissioning Plan which include the Integrated concept of Decontamination and Decommissioning and Radwaste Management, reflects the substantial work that has been incorporated by IFIN-HH in collaboration with SITON, which has resulted in substantial improvement in document The decommissioning strategy must take into account costs for VVR-S Reactor decommissioning, as well as costs for much needed refurbishments to the radioactive waste treatment plant and the Baita-Bihor waste disposal repository. Several improvements to the Baita-Bihor repository and IFIN-HH waste treatment facility were proposed. The quantities and composition of the radioactive waste generated by VVR-S Reactor dismantling were again estimated by streams and the best demonstrated practicable processing solution was proposed. The estimated quantities of materials to be managed in the near future raise some issues that need to be solved swiftly, such as treatment of aluminum and lead and graphite management. It is envisaged that these materials to be treated to Subsidiary for Nuclear Research (SCN) Pitesti. (authors)

Barariu, G. [National Authority for Nuclear Activity-Subsidiary of Technology and Engineering for Nuclear Projects (Romania)

2008-07-01T23:59:59.000Z

369

Shipper/receiver difference verification of spent fuel by use of PDET  

Science Conference Proceedings (OSTI)

Spent fuel storage pools in most countries are rapidly approaching their design limits with the discharge of over 10,000 metric tons of heavy metal from global reactors. Countries like UK, France or Japan have adopted a closed fuel cycle by reprocessing spent fuel and recycling MOX fuel while many other countries opted for above ground interim dry storage for their spent fuel management strategy. Some countries like Finland and Sweden are already well on the way to setting up a conditioning plant and a deep geological repository for spent fuel. For all these situations, shipments of spent fuel are needed and the number of these shipments is expected to increase significantly. Although shipper/receiver difference (SRD) verification measurements are needed by IAEA when the recipient facility receives spent fuel, these are not being practiced to the level that IAEA has desired due to lack of a credible measurement methodology and instrument that can reliably perform these measurements to verify non-diversion of spent fuel during shipment and confirm facility operator declarations on the spent fuel. In this paper, we describe a new safeguards method and an associated instrument, Partial Defect Tester (PDET), which can detect pin diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies in an in-situ condition. The PDET uses multiple tiny neutron and gamma detectors in the form of a cluster and a simple, yet highly precise, gravity-driven system to obtain underwater radiation measurements inside a Pressurized Water Reactor (PWR) spent fuel assembly. The method takes advantage of the PWR fuel design which contains multiple guide tubes which can be accessed from the top. The data obtained in such a manner can provide spatial distribution of neutron and gamma flux within a spent fuel assembly. Our simulation study as well as validation measurements indicated that the ratio of the gamma signal to the thermal neutron signal at each detector location normalized to the peak ratio of all the detector locations gives a unique signature that is sensitive to missing pins. The signature is principally dependent on the geometry of the detector locations, and little sensitive to enrichment or burn-up variations. A small variation in the fuel bundle, such as a few missing pins, changes the shape of the signature to enable detection. After verification of the non-diversion of spent fuel pins, the neutron signal and gamma signal are subsequently used to verify the consistency of the operator declaration on the fuel burn-up and cooling time. (authors)

Ham, Y. S.; Sitaraman, S. [Global Security Directorate, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States)

2011-07-01T23:59:59.000Z

370

Naval Spent Fuel Rail Shipment Accident Exercise Objectives ...  

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

Naval Spent Fuel Rail Shipment Accident Exercise Objectives Naval Spent Fuel Rail Shipment Accident Exercise Objectives Naval Spent Fuel Rail Shipment Accident Exercise Objectives...

371

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

372

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

Science Conference Proceedings (OSTI)

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

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

1980-05-01T23:59:59.000Z

373

An Evaluation of the Concept of Transuranic Burning Using Liquid Metal Reactors  

Science Conference Proceedings (OSTI)

This evaluation investigates the potential benefits of separating the transuranic elements from spent reactor fuel before it is disposed of in geologic repositories. It addresses the question: Would the benefits to radioactive waste disposal justify both processing the spent fuel and deploying liquid metal reactors (LMRs) to transmute the separated transuranics?

1991-03-25T23:59:59.000Z

374

Environmental Statements, Availability, Etc., Programmatic Spent...  

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

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

375

HTGR Spent Fuel Treatment Program. HTGR Spent Fuel Treatment Development Program Plan  

SciTech Connect

The spent fuel treatment (SFT) program plan addresses spent fuel volume reduction, packaging, storage, transportation, fuel recovery, and disposal to meet the needs of the HTGR Lead Plant and follow-on plants. In the near term, fuel refabrication will be addressed by following developments in fresh fuel fabrication and will be developed in the long term as decisions on the alternatives dictate. The formulation of this revised program plan considered the implications of the Nuclear Waste Policy Act of 1982 (NWPA) which, for the first time, established a definitive national policy for management and disposal of nuclear wastes. Although the primary intent of the program is to address technical issues, the divergence between commercial and government interests, which arises as a result of certain provisions of the NWPA, must be addressed in the economic assessment of technically feasible alternative paths in the management of spent HTGR fuel and waste. This new SFT program plan also incorporates a significant cooperative research and development program between the United States and the Federal Republic of Germany. The major objective of this international program is to reduce costs by avoiding duplicate efforts.

1984-12-01T23:59:59.000Z

376

Monte Carlo methods of neutron beam design for neutron capture therapy at the MIT Research Reactor (MITR-II)  

Science Conference Proceedings (OSTI)

Monte Carlo methods of coupled neutron/photon transport are being used in the design of filtered beams for Neutron Capture Therapy (NCT). This method of beam analysis provides segregation of each individual dose component, and thereby facilitates beam optimization. The Monte Carlo method is discussed in some detail in relation to NCT epithermal beam design. Ideal neutron beams (i.e., plane-wave monoenergetic neutron beams with no primary gamma-ray contamination) have been modeled both for comparison and to establish target conditions for a practical NCT epithermal beam design. Detailed models of the 5 MWt Massachusetts Institute of Technology Research Reactor (MITR-II) together with a polyethylene head phantom have been used to characterize approximately 100 beam filter and moderator configurations. Using the Monte Carlo methodology of beam design and benchmarking/calibrating our computations with measurements, has resulted in an epithermal beam design which is useful for therapy of deep-seated brain tumors. This beam is predicted to be capable of delivering a dose of 2000 RBE-cGy (cJ/kg) to a therapeutic advantage depth of 5.7 cm in polyethylene assuming 30 micrograms/g 10B in tumor with a ten-to-one tumor-to-blood ratio, and a beam diameter of 18.4 cm. The advantage ratio (AR) is predicted to be 2.2 with a total irradiation time of approximately 80 minutes. Further optimization work on the MITR-II epithermal beams is expected to improve the available beams. 20 references.

Clement, S.D.; Choi, J.R.; Zamenhof, R.G.; Yanch, J.C.; Harling, O.K. (Massachusetts Institute of Technology, Cambridge (USA))

1990-01-01T23:59:59.000Z

377

Developing fuel management capabilities based on coupled Monte Carlo depletion in support of the MIT Research Reactor (MITR) conversion  

E-Print Network (OSTI)

Pursuant to a 1986 NRC ruling, the MIT Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. Prior studies have shown that the MITR will be able to ...

Romano, Paul K. (Paul Kollath)

2009-01-01T23:59:59.000Z

378

Mission Need Statement: Idaho Spent Fuel Facility Project  

SciTech Connect

Approval is requested based on the information in this Mission Need Statement for The Department of Energy, Idaho Operations Office (DOE-ID) to develop a project in support of the mission established by the Office of Environmental Management to "complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research". DOE-ID requests approval to develop the Idaho Spent Fuel Facility Project that is required to implement the Department of Energy's decision for final disposition of spent nuclear fuel in the Geologic Repository at Yucca Mountain. The capability that is required to prepare Spent Nuclear Fuel for transportation and disposal outside the State of Idaho includes characterization, conditioning, packaging, onsite interim storage, and shipping cask loading to complete shipments by January 1,2035. These capabilities do not currently exist in Idaho.

Barbara Beller

2007-09-01T23:59:59.000Z

379

Mission Need Statement: Idaho Spent Fuel Facility Project  

SciTech Connect

Approval is requested based on the information in this Mission Need Statement for The Department of Energy, Idaho Operations Office (DOE-ID) to develop a project in support of the mission established by the Office of Environmental Management to "complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research". DOE-ID requests approval to develop the Idaho Spent Fuel Facility Project that is required to implement the Department of Energy's decision for final disposition of spent nuclear fuel in the Geologic Repository at Yucca Mountain. The capability that is required to prepare Spent Nuclear Fuel for transportation and disposal outside the State of Idaho includes characterization, conditioning, packaging, onsite interim storage, and shipping cask loading to complete shipments by January 1,2035. These capabilities do not currently exist in Idaho.

Barbara Beller

2007-09-01T23:59:59.000Z

380

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

Note: This page contains sample records for the topic "research reactor spent" 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

Advanced burner test reactor preconceptual design report.  

Science Conference Proceedings (OSTI)

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

2008-12-16T23:59:59.000Z

382

Biosciences Division: Endurance Bioenergy Reactor(tm)  

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

Endurance Bioenergy Reactor(tm) DOE Logo Search BIO ... Search Argonne Home > BIO home > Endurance Bioenergy Reactor(tm) BIO Home Page About BIO News Releases Research Publications...

383

RERTR program activities related to the development and application of new LEU fuels. [Reduced Enrichment Research and Test Reactor; low-enriched uranium  

SciTech Connect

The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm/sup 3/ each year, from the current 1.7 g U/cm/sup 3/ to the 7.0 g U/cm/sup 3/ which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years.

Travelli, A.

1983-01-01T23:59:59.000Z

384

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

SciTech Connect

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

Trond Bjornard; Philip C. Durst

2012-05-01T23:59:59.000Z

385

Report, Long-Term Nuclear Technology Research and Development Plan |  

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

Report, Long-Term Nuclear Technology Research and Development Plan Report, Long-Term Nuclear Technology Research and Development Plan Report, Long-Term Nuclear Technology Research and Development Plan This document constitutes the first edition of a long-term research and development (R&D) plan for nuclear technology in the United States. The federally-sponsored nuclear technology programs of the United States are almost exclusively the province of the U.S. Department of Energy (DOE). The nuclear energy areas in DOE include, but are not limited to, R&D related to power reactors and the responsibility for the waste management system for final disposition of the spent fuel resulting from nuclear power reactors. Although a major use of nuclear technology is to supply energy for electricity production, the DOE has far broader roles regarding nuclear

386

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

387

Characterization of spent fuel approved testing material: ATM-106  

Science Conference Proceedings (OSTI)

The characterization data obtained to date are described for Approved Testing Material (ATM)-106 spent fuel from Assembly BT03 of pressurized-water reactor Calvert Cliffs No. 1. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well- characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCWRM) program. ATM-106 consists of 20 full-length irradiated fuel rods with rod-average burnups of about 3700 GJ/kgM (43 MWd/kgM) and expected fission gas release of /approximately/10%. Characterization data include (1) as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) calculated nuclide inventories and radioactivities in the fuel and cladding; and (6) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel rod are being conducted and will be included in planned revisions of this report. 12 refs., 110 figs., 81 tabs.

Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thornhill, C.K.

1988-10-01T23:59:59.000Z

388

Characterization of spent fuel approved testing material---ATM-105  

Science Conference Proceedings (OSTI)

The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report.

Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

1991-12-01T23:59:59.000Z

389

Characterization of spent fuel approved testing material--ATM-104  

SciTech Connect

The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission products. Characterization data include (1) fabricated fuel design, irradiation history, and subsequent storage and handling history; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM) and electron probe microanalyses (EPMA); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding.

Guenther, R.J.; Blahnik, D.E.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

1991-12-01T23:59:59.000Z

390

Utilization of MIT research reactor by Boston-area universities. Final report, July 1, 1978-June 30, 1980  

SciTech Connect

The guest institutions which used the MITR reactor during 1978-1979 are listed. The participation during previous years since the program began at MIT is also indicated. A summary of the type and amount of participation by local institutions is given.

Clark, L. Jr.; Janghorbani, M.

1981-07-01T23:59:59.000Z

391

Plutonium recovery from spent reactor fuel by uranium displacement  

DOE Patents (OSTI)

A process for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

Ackerman, John P. (Downers Grove, IL)

1992-01-01T23:59:59.000Z

392

Plutonium recovery from spent reactor fuel by uranium displacement  

DOE Patents (OSTI)

This report discusses a process for separating uranium values and transuranic values from fission products containing rare earth values when the values which are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is re-established.

Ackerman, J.P.

1991-01-01T23:59:59.000Z

393

Establishing a Cost Basis for Converting the High Flux Isotope Reactor from High Enriched to Low Enriched Uranium Fuel  

Science Conference Proceedings (OSTI)

Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration /Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.

Primm, Trent [ORNL; Guida, Tracey [University of Pittsburgh

2010-02-01T23:59:59.000Z

394

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

395

Summary report on the HFED (High-Uranium-Loaded Fuel Element Development) miniplate irradiations for the RERTR (Reduced Enrichment Research and Test Reactor) Program  

SciTech Connect

An experiment to evaluate the irradiation characteristics of various candidate low-enriched, high-uranium content fuels for research and test reactors was performed for the US Department of Energy Reduced Enrichment Research and Test Reactor Program. The experiment included the irradiation of 244 miniature fuel plates (miniplates) in a core position in the Oak Ridge Research Reactor. The miniplates were aluminum-based, dispersion-type plates 114.3 mm long by 50.8 mm wide with overall plate thicknesses of 1.27 or 1.52 mm. Fuel core dimensions varied according to the overall plate thicknesses with a minimum clad thickness of 0.20 mm. Tested fuels included UAl/sub x/, UAl/sub 2/, U/sub 3/O/sub 8/, U/sub 3/SiAl, U/sub 3/Si, U/sub 3/Si/sub 1.5/, U/sub 3/Si/sub 2/, U/sub 3/SiCu, USi, U/sub 6/Fe, and U/sub 6/Mn/sub 1.3/ materials. Although most miniplates were made with low-enriched uranium (19.9%), some with medium-enriched uranium (40 to 45%), a few with high-enriched uranium (93%), and a few with depleted uranium (0.2 to 0.4%) were tested for comparison. These fuel materials were irradiated to burnups ranging from /approximately/27 to 98 at. % /sup 235/U depletion. Operation of the experiment, measurement of miniplate thickness as the irradiation progressed, ultimate shipment of the irradiated miniplates to various hot cells, and preliminary results are reported here. 18 refs., 12 figs., 7 tabs.

Senn, R.L.

1989-04-01T23:59:59.000Z

396

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

397

The Innovations, Technology and Waste Management Approaches to Safely Package and Transport the World's First Radioactive Fusion Research Reactor for Burial  

SciTech Connect

Original estimates stated that the amount of radioactive waste that will be generated during the dismantling of the Tokamak Fusion Test Reactor will approach two million kilograms with an associated volume of 2,500 cubic meters. The materials were activated by 14 MeV neutrons and were highly contaminated with tritium, which present unique challenges to maintain integrity during packaging and transportation. In addition, the majority of this material is stainless steel and copper structural metal that were specifically designed and manufactured for this one-of-a-kind fusion research reactor. This provided further complexity in planning and managing the waste. We will discuss the engineering concepts, innovative practices, and technologies that were utilized to size reduce, stabilize, and package the many unique and complex components of this reactor. This waste was packaged and shipped in many different configurations and methods according to the transportation regulations and disposal facility requirements. For this particular project, we were able to utilize two separate disposal facilities for burial. This paper will conclude with a complete summary of the actual results of the waste management costs, volumes, and best practices that were developed from this groundbreaking and successful project.

Keith Rule; Erik Perry; Jim Chrzanowski; Mike Viola; Ron Strykowsky

2003-09-15T23:59:59.000Z

398

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

399

Sandia National Laboratories: Research: Facilities: Annular Core...  

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

Annular Core Research Reactor facility Nuclear science photo At the Annular Core Research Reactor (ACRR) facility, Sandia researchers can subject various test objects to a mixed...

400

Reactor operations informal report, October 1994  

Science Conference Proceedings (OSTI)

This monthly progress report is divided into two parts. Part one covers the Brookhaven Medical Research Reactor and part two covers the Brookhaven High Flux Beam Reactor. Information is given for each reactor covering the following areas: reactor operation; instrumentation; mechanical maintenance; occurrence reports; and reactor safety.

Hauptman, H.M.; Petro, J.N.; Jacobi, O.; Lettieri, V.; Holden, N.; Ports, D.; Petricek, R.

1994-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "research reactor spent" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

Behavior of spent fuel under unsaturated conditions  

Science Conference Proceedings (OSTI)

To evaluate the performance of spent fuel in the potential repository at Yucca Mountain, Nevada, spent fuel fragments are being exposed to small and intermittent amounts of simulated groundwater under unsaturated conditions. Both the leachate and the visual appearance of the spent fuel have been characterized for 581 days of testing. The amount of Am and Cm measured in the leachates was one to two orders of magnitude greater than that released from spent fuel under saturated conditions. The cause of this difference has not been firmly identified but may be attributable to the presence of large amounts of actinide-containing colloids in the leachate of the unsaturated tests.

Finn, P.A.; Hoh, J.C.; Bates, J.K.; Wolf, S.F. [Argonne National Lab., IL (United States). Chemical Technology Div.

1994-12-31T23:59:59.000Z

402

HYDRAULIC CEMENT PREPARATION FROM LURGI SPENT SHALE  

E-Print Network (OSTI)

P. , "Investigations on hydraulic cement from spent oilCO, April 16-18, 1980 HYDRAULIC CEMENT PREPARATION FROMUniversity of California. HYDRAULIC CEMENT PREPARATION FROM

Mehta, P.K.

2013-01-01T23:59:59.000Z

403

HYDRAULIC CEMENT PREPARATION FROM LURGI SPENT SHALE  

E-Print Network (OSTI)

showing potential for subsidence and spent shale leaching.cracking and ground subsidence, and low leaving largeto 210 m overburden), and subsidence. These problems may be

Mehta, P.K.

2013-01-01T23:59:59.000Z

404

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

405

Effect of changes in DOE pricing policies for enrichment and reprocessing on research reactor fuel cycle costs  

SciTech Connect

Fuel cycle costs with HEU and LEU fuels for the IAEA generic 10 MW reactor are updated to reflect the change in DOE pricing policy for enrichment services as of October 1985 and the published charges for LEU reprocessing services as of February 1986. The net effects are essentially no change in HEU fuel cycle costs and a reduction of about 8 to 10% in the fuel cycle costs for LEU silicide fuel.

Matos, J.E.; Freese, K.E.

1986-11-03T23:59:59.000Z

406

Research into the pyrolysis of pure cellulose, lignin, and birch wood flour in the China Lake entrained-flow reactor  

DOE Green Energy (OSTI)

This experimental program used the China Lake entrained-flow pyrolysis reactor to briefly investigate the pyrolysis of pure cellulose, pure lignin, and birch wood flour. The study determined that the cellulose and wood flour do pyrolyze to produce primarily gaseous products containing significant amounts of ethylene and other useful hydrocarbons. During attempts to pyrolyze powdered lignin, the material melted and bubbled to block the reactor entrance. The pure cellulose and wood flour produced C/sub 2/ + yields of 12% to 14% by weight, which were less than yields from an organic feedstock derived from processed municipal trash. The char yields were 0.1% by weight from cellulose and 1.5% from birch wood flour - one to two orders of magnitude less than were produced from the trash-derived feedstock. In scanning electron microscope photographs, most of the wood flour char had a sintered and agglomerated appearance, although some particles retained the gross cell characteristics of the wood flour. The appearance of the char particles indicated that the material had once been molten and possibly vapor before it formed spheroidal particles about 1 ..mu..m diameter which agglomerated to form larger char particles. The ability to completely melt or vaporize lignocellulosic materials under conditions of high heating rates has now been demonstrated in a continuous flow reactor and promises new techniques for fast pyrolysis. This char was unexpectedly attracted by a magnet, presumably because of iron contamination from the pyrolysis reactor tube wall. The production of water-insoluble tars was negligible compared to the tars produced from trash-derived feedstock. The production of water-soluble organic materials was fairly low and qualitatively appeared to vary inversely with temperature. This study was of a preliminary nature and additional studies are necessary to optimize ethylene production from these feedstocks.

Diebold, J.

1980-06-01T23:59:59.000Z

407

Italian hybrid and fission reactors scenario analysis  

SciTech Connect

Italy is a country where a long tradition of studies both in the fission and fusion field is consolidated; nevertheless a strong public opinion concerned with the destination of the Spent Nuclear Fuel hinders the development of nuclear power. The possibility to a severe reduction of the NSF mass generated from a fleet of nuclear reactors employing an hypothetical fusionfission hybrid reactor has been investigated in the Italian framework. The possibility to produce nuclear fuel for the fission nuclear reactors with the hybrid reactor was analyzed too.

Ciotti, M.; Manzano, J.; Sepielli, M. [ENEA CR Frascati, Via Enrico Fermi, 45, 00044, Frascati, Roma (Italy); ENEA CR casaccia, Via Anguillarese, 301, 00123, Santa Maria di Galeria, Roma (Italy)

2012-06-19T23:59:59.000Z

408

Accelerator-driven transmutation of spent fuel elements  

DOE Patents (OSTI)

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

Venneri, Francesco (Los Alamos, NM); Williamson, Mark A. (Los Alamos, NM); Li, Ning (Los Alamos, NM)

2002-01-01T23:59:59.000Z

409

Characteristics of fuel crud and its impact on storage, handling, and shipment of spent fuel. [Fuel crud  

Science Conference Proceedings (OSTI)

Corrosion products, called ''crud,'' form on out-of-reactor surfaces of nuclear reactor systems and are transported by reactor coolant to the core, where they deposit on external fuel-rod cladding surfaces and are activated by nuclear reactions. After discharge of spent fuel from a reactor, spallation of radioactive crud from the fuel rods could impact wet or dry storage operations, handling (including rod consolidation), and shipping. It is the purpose of this report to review earlier (1970s) and more recent (1980s) literature relating to crud, its characteristics, and any impact it has had on actual operations. Crud characteristics vary from reactor type to reactor