Powered by Deep Web Technologies
Note: This page contains sample records for the topic "reactor aging management" 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.


1

International Forum for Reactor Aging Management (IFRAM)  

SciTech Connect

The Nuclear Regulatory Commission has undertaken a program to lay the groundwork for defining proactive actions to manage degradation of materials in light water reactors (LWRs). This article discusses the international forum for reactor aging management.

Bond, Leonard J.

2010-11-01T23:59:59.000Z

2

Materials Reliability Program: Pressurized Water Reactor Internals Aging Management Program Development Template (MRP-342)  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) has completed and published guidance for managing the effects of aging degradation in pressurized water reactor (PWR) internals. The initial version of this report, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 0), was submitted to the staff of the U. S. Nuclear Regulatory Commission (NRC) ...

2012-10-23T23:59:59.000Z

3

Developing PWR Aging-Management Strategies for Reactor Vessel ...  

Science Conference Proceedings (OSTI)

AREVA Fuel Condition Index for a Pressurized Water Reactor .... Stress Corrosion Cracking Behavior near the Fusion Boundary of Dissimilar Weld Joint with ...

4

Semi-Annual Report on Work Supporting the International Forum for Reactor Aging Management (IFRAM)  

SciTech Connect

During the first six months of this project, Pacific Northwest National Laboratory has provided planning and leadership support for the establishment of the International Forum for Reactor Aging Management (IFRAM). This entailed facilitating the efforts of the Global Steering Committee to prepare the charter, operating guidelines, and other documents for IFRAM. It also included making plans for the Inaugural meeting and facilitating its success. This meeting was held on August 4 5, 2011, in Colorado Springs, Colorado. Representatives from Asia, Europe, and the United States met to share information on reactor aging management and to make plans for the future. Professor Tetsuo Shoji was elected chairperson of the Leadership Council. This kick-off event transformed the dream of an international forum into a reality. On August 4-5, 2011, IFRAM began to achieve its mission. The work completed successfully during this period was built upon important previous efforts. This included the development of a proposal for establishing IFRAM and engaging experts in Asia and Europe. The proposal was presented at Engagement workshops in Seoul, Korea (October 2009) and Petten, The Netherlands (May 2010). Participants in both groups demonstrated strong interest in the establishment of IFRAM. Therefore, the Global Steering Committee was formed to plan and carry out the start-up of IFRAM in 2011. This report builds on the initial activities and documents the results of activities over the last six months.

Bond, Leonard J.; Brenchley, David L.

2011-11-30T23:59:59.000Z

5

Proceedings of the Inaugural Meeting of the International Forum for Reactor Aging Management (IFRAM)  

SciTech Connect

In almost all countries with nuclear power plants (NPPs), regulatory authorities and the nuclear industry are looking at some form of extended operating periods. To support life extension activities it is necessary to ensure the continued safety and reliability of system, structures, and components, and the component materials. Internationally, a variety of individual national and international activities have been initiated including Plant Life Management through the International Atomic Energy Agency, Electric Power Research Institute’s Long Term Operation program, and various national programs in managing materials degradation and related topics. The U.S. Nuclear Regulatory Commission (NRC) engaged the international community in workshops in 2005-2006 to identify research needs and to collect information in an expert panel report on Proactive Management of Materials Degradation (PMMD), which was reported in NUREG/CR-6923. These results are also available via an Information Tool on the internet at http://pmmd.pnl.gov. This information builds on the extensive compilations known as the GALL Report (Generic Aging Lessons Learned, NUREG-1801, Vols. 1 and 2). Pacific Northwest National Laboratory (PNNL) recently issued a report on the review of various international activities in PMMD (PNNL-17779). There have also been initiatives by Electricite de France, Tokyo Electric Power Company, EPRI, and others to establish a "Materials Aging Institute." Within the materials degradation research community there are also networks and technical meetings focused on some elements of PMMD. In spite of all these efforts, there is currently no forum to bring together these diverse activities and provide coordinated information exchange and prioritization of materials aging management/PMMD topics. It is believed that the International Forum for Reactor Aging Management (IFRAM) would be a good way to achieve this goal and help develop new approaches for ensuring continued safe operation in existing and future nuclear power plants. To begin addressing this need, NRC has established a Proactive Management of Materials Degradation Program for managing in-service degradation of metallic components in aging NPPs. The NRC is seeking to facilitate the establishment of IFRAM as a network of international experts who would exchange information on operating experience, best practices, and emerging knowledge. These experts would be willing to work jointly and leverage the separate efforts of existing national programs into a unified approach to enable the potential for the safe and economic life extension of NPPs. A proposal for establishing IFRAM was developed and presented at Engagement workshops in Seoul, Korea (October 2009) and Petten, The Netherlands (May 2010). Participants in both groups demonstrated strong interest in the establishment of IFRAM. Therefore the Global Steering Committee was formed to plan and carry out the start-up of IFRAM in 2011. This group finalized the documents for IFRAM and organized the kick-off meeting. This document records the contents of the inaugural meeting of IFRAM, which was held August 4-5, 2011, at the Cheyenne Mountain Resort in Colorado Springs, Colorado. Representatives from Asia, Europe, and the United States met to share information on reactor aging management and make plans for the future. Professor Tetsuo Shoji was elected chairperson of the Leadership Council. This kick-off event transformed the dream of an international forum into a reality. IFRAM has begun to achieve its mission.

Bond, Leonard J.; Brenchley, David L.

2011-09-01T23:59:59.000Z

6

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

7

Management of the aging of critical safety-related concrete structures in light-water reactor plants  

SciTech Connect

The Structural Aging Program has the overall objective of providing the USNRC with an improved basis for evaluating nuclear power plant safety-related structures for continued service. The program consists of a management task and three technical tasks: materials property data base, structural component assessment/repair technology, and quantitative methodology for continued-service determinations. Objectives, accomplishments, and planned activities under each of these tasks are presented. Major program accomplishments include development of a materials property data base for structural materials as well as an aging assessment methodology for concrete structures in nuclear power plants. Furthermore, a review and assessment of inservice inspection techniques for concrete materials and structures has been complete, and work on development of a methodology which can be used for performing current as well as reliability-based future condition assessment of concrete structures is well under way. 43 refs., 3 tabs.

Naus, D.J.; Oland, C.B. (Oak Ridge National Lab., TN (USA)); Arndt, E.G. (Nuclear Regulatory Commission, Washington, DC (USA))

1990-01-01T23:59:59.000Z

8

Light Water Reactor Sustainability Program: Materials Aging and Degradation  

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

Materials Aging and Materials Aging and Degradation Technical Program Plan Light Water Reactor Sustainability Program: Materials Aging and Degradation Technical Program Plan Components serving in a nuclear reactor plant must withstand a very harsh environment including extended time at temperature, neutron irradiation, stress, and/or corrosive media. The many modes of degradation are complex and vary depending on location and material. However, understanding and managing materials degradation is a key for the continued safe and reliable operation of nuclear power plants. Extending reactor service to beyond 60 years will increase the demands on materials and components. Therefore, an early evaluation of the possible effects of extended lifetime is critical. The recent NUREG/CR-6923 gives a

9

Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals (MRP-232)  

Science Conference Proceedings (OSTI)

This report summarizes the aging management strategy development for Westinghouse and Combustion Engineering (CE) reactor internals. This report provides the technical basis for the aging management requirements of Westinghouse and CE reactor internals in the Pressurized Water Reactor (PWR) internals I&E guidelines (MRP-227-Rev. 0).

2008-12-22T23:59:59.000Z

10

Control Relay Aging Management Guideline  

Science Conference Proceedings (OSTI)

The concept for this report came from the concern that many control relays have been in service for an extended period of time and an effective aging management program may not be in place for these relays. In addition, recent Institute of Nuclear Power Operations (INPO) data indicate that relays are one of the leading component types causing scrams. With control relay age increasing and relays being a significant contributor to scrams, an evaluation of control relay maintenance and replacement strategie...

2011-05-03T23:59:59.000Z

11

Materials Reliability Program: Strategies for Managing Aging Effects in PWR Vessel Internals - Interim Update (MRP-99)  

Science Conference Proceedings (OSTI)

This report updates the previous EPRI report on developing strategies for managing aging effects in pressurized water reactor (PWR) internals during the license renewal term.

2003-12-04T23:59:59.000Z

12

New Fuel Cycle and Fuel Management Options in Heavy Liquid Metal-Cooled Reactors  

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Fuel Management of Reactors Other Than Light Water Reactors

Ehud Greenspan; Pavel Hejzlar; Hiroshi Sekimoto; Georgy Toshinsky; David Wade

13

Reactor Vessel Embrittlement Management Handbook: A Handbook for Managing Reactor Vessel Embrittlement and Vessel Integrity  

Science Conference Proceedings (OSTI)

For many reactor pressure vessels, embrittlement is the primary concern for continued safe operation. The shutdown of the Yankee Rowe plant because of uncertainties related to embrittlement of the vessel demonstrates the importance of adequately addressing embrittlement issues. Managing embrittlement requires integration, management, and implementation of diverse technical, regulatory, planning, and economic activities. An effective embrittlement management program will ensure vessel safety and reliabili...

1994-01-22T23:59:59.000Z

14

Low-Voltage Cable Aging Management Guidance  

Science Conference Proceedings (OSTI)

This report, intended for use by power plant engineers and craft and administrative personnel who are responsible for cable system aging management, provides guidance on the effective and efficient application of current information and techniques related to aging and condition monitoring (CM) of low-voltage cables. The identification of cables that are prone to early aging and the timely replacement or repair of cables that are degraded will minimize unnecessary power losses, unplanned shutdowns, ...

2013-12-23T23:59:59.000Z

15

Materials Reliability Program: San Onofre Nuclear Generating Station Reactor Vessel Internals Management Engineering Program (MRP-303)  

Science Conference Proceedings (OSTI)

All operating pressurized water reactors must have a reactor vessel internals aging management document in place by December 2011 according to the mandatory requirement under Nuclear Energy Institute (NEI) 03-08. This program should be developed to meet the guidance provided by Materials Reliability Program (MRP) -227, Rev. 0, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines. For non-license renewal plants, the requirements are valid within the current license period, and the Elec...

2011-02-28T23:59:59.000Z

16

Risk management activities at the DOE Class A reactor facilities  

Science Conference Proceedings (OSTI)

The probabilistic risk assessment (PRA) and risk management group of the Association for Excellence in Reactor Operation (AERO) develops risk management initiatives and standards to improve operation and increase safety of the DOE Class A reactor facilities. Principal risk management applications that have been implemented at each facility are reviewed. The status of a program to develop guidelines for risk management programs at reactor facilities is presented.

Sharp, D.A. [Westinghouse Savannah River Co., Aiken, SC (United States); Hill, D.J. [Argonne National Lab., IL (United States); Linn, M.A. [Oak Ridge National Lab., TN (United States); Atkinson, S.A. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Hu, J.P. [Brookhaven National Lab., Upton, NY (United States)

1993-12-31T23:59:59.000Z

17

Aging Management Guideline for commercial nuclear power plants: Motor control centers; Final report  

SciTech Connect

This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) commercial nuclear power plant motor control centers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

Toman, G.; Gazdzinski, R.; O`Hearn, E. [Ogden Environmental and Energy Services Co., Inc., Blue Bell, PA (United States)

1994-02-01T23:59:59.000Z

18

Managing Aging Effects on Dry Cask Storage  

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

Managing Aging Effects Managing Aging Effects on Dry Cask Storage Systems for Extended Long-Term Storage and Transportation of Used Fuel Rev. 0 Prepared for U.S. Department of Energy Used Fuel Disposition Campaign O.K. Chopra, D. Diercks, R. Fabian, D. Ma, V. Shah, S-W Tam, and Y.Y. Liu Argonne National Laboratory June 30, 2012 FCRD-USED-2012-000119 ANL-12/29 DISCLAIMER This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness, of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately

19

Management of Naval Reactors' Cyber Security Program, OIG-0884  

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

Naval Reactors' Naval Reactors' Cyber Security Program DOE/IG-0884 April 2013 U.S. Department of Energy Office of Inspector General Office of Audits and Inspections Department of Energy Washington, DC 20585 April 12, 2013 MEMORANDUM FOR THE SECRETARY FROM: Gregory H. Friedman Inspector General SUBJECT: INFORMATION: Audit Report on "Management of Naval Reactors' Cyber Security Program" INTRODUCTION AND OBJECTIVE The Naval Reactors Program (Naval Reactors), an organization within the National Nuclear Security Administration, provides the military with safe and reliable nuclear propulsion plants to power warships and submarines. Naval Reactors maintains responsibility for activities supporting the United States Naval fleet nuclear propulsion systems, including research and

20

CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor |  

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

Management- Oak Ridge National Laboratory High Flux Isotope Management- Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Management in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope

Note: This page contains sample records for the topic "reactor aging management" 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

DETERMINING THE EFFECTS OF RADIATION ON AGING CONCRETE STRUCTURES OF NUCLEAR REACTORS  

SciTech Connect

The U.S. Department of Energy Office of Environmental Management (DOE-EM) is responsible for the Decontamination and Decommissioning (D&D) of nuclear facilities throughout the DOE Complex. Some of these facilities will be completely dismantled, while others will be partially dismantled and the remaining structure will be stabilized with cementitious fill materials. The latter is a process known as In-Situ Decommissioning (ISD). The ISD decision process requires a detailed understanding of the existing facility conditions, and operational history. System information and material properties are need for aged nuclear facilities. This literature review investigated the properties of aged concrete structures affected by radiation. In particular, this review addresses the Savannah River Site (SRS) isotope production nuclear reactors. The concrete in the reactors at SRS was not seriously damaged by the levels of radiation exposure. Loss of composite compressive strength was the most common effect of radiation induced damage documented at nuclear power plants.

Serrato, M.

2010-01-29T23:59:59.000Z

22

Managing Aging Effects on Dry Cask Storage Systems for Extended...  

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

Managing Aging Effects on Dry Cask Storage Systems for Extended Long Term Storage and Transportation of Used Fuel Rev0 Managing Aging Effects on Dry Cask Storage Systems for...

23

Boiling-Water Reactor internals aging degradation study. Phase 1  

SciTech Connect

This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.

Luk, K.H. [Oak Ridge National Lab., TN (United States)

1993-09-01T23:59:59.000Z

24

Chapter 5, "License Renewal and Aging Management for Continued Service  

SciTech Connect

As of August 2011, there were 104 commercial nuclear power reactors licensed to operate in 31 states in the United States. Initial operating licenses in the United States are granted for a period of 40 years. In order to help assure an adequate energy supply, the USNRC has established a timely license renewal process and clear requirements that are needed to ensure safe plant operation for an extended plant life. The principals of license renewal and the basic requirements that address license renewal are identified as well as additional sources of guidance that can be utilized as part of the license renewal process. Aging management program inspections and operating experience related to the concrete and steel containment structures are provided. Finally, several lessons learned are provided based on containment operating experience.

Naus, Dan J [ORNL] [ORNL

2013-01-01T23:59:59.000Z

25

CRAD, Management - Oak Ridge National Laboratory High Flux Isotope Reactor  

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

Oak Ridge National Laboratory High Flux Isotope Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Management portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR More Documents & Publications

26

Plant Engineering: Cable Aging Management Program Implementation Guidance  

Science Conference Proceedings (OSTI)

This report describes issues identified during the initial implementation of low- and medium-voltage cable aging management programs by utilities and presents possible resolutions and guidance.

2011-12-01T23:59:59.000Z

27

Effectiveness of storage practices in mitigating aging degradation during reactor layup  

Science Conference Proceedings (OSTI)

One of the issues identified in the US Nuclear Regulatory Commission`s Nuclear Plant Aging Research program plan is the need to understand the state of ``mothballed`` or other out-of-service equipment to ensure subsequent safe operation. Programs for proper storage and preservation of materials and components are required by NRC regulations (10 CFR 50, Appendix B). However, materials and components have been seriously degraded due to improper storage, protection, or layup, at facilities under construction as well as those with operating licenses. Pacific Northwest Laboratory has evaluated management of aging for unstarted or mothballed nuclear power plants. The investigations revealed that no uniform guidance in the industry addresses reactor layup. In each case investigated, layup was not initiated in a timely manner, primarily because of schedule uncertainty. Hence, it is reasonable to assume that this delay resulted in accelerated aging of some safety-significant structures, systems, and components (SSCs). The applicable layup process is site-specific. The reactor type, climatic setting, operational status, and materials of construction are factors that strongly dictate the layup method to be used. The adequacy of current layup practices, and hence their impact on safety-significant SSCS, is not fully understood.

Enderlin, W.I.

1995-09-01T23:59:59.000Z

28

Impact of lattice geometry distortion due to ageing on selected physics parameters of a CANDU reactor  

SciTech Connect

In this paper, results related to a limited scope assessment of the geometry-distortion-induced effects on key reactor physics parameters of a CANDU reactor are discussed. These results were generated by simulations using refined analytical methods and detailed modeling of CANDU reactor core with aged lattice cell geometry. (authors)

Tholammakkil, J. [Canadian Nuclear Safety Commission, P.O. Box 1046, Ottawa, Ont. (Canada); Rahnema, F. [Georgia Inst. of Technology, Atlanta, GA 30332-0405 (United States); Ilas, G. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Stamm'ler, R. [Studsvik Scandpower AS, P.O. Box 15, NO-2027 Kjeller (Norway); Khotylev, V.; Serghiuta, D. [Canadian Nuclear Safety Commission, P.O. Box 1046, Ottawa, Ont. (Canada)

2006-07-01T23:59:59.000Z

29

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

30

Cable Aging Management Training Course for Low-Voltage Cables  

Science Conference Proceedings (OSTI)

The "EPRI PSE Cable Aging Training Course for Low-Voltage Cables" provides a basis for understanding of the aging of the insulation and jackets of electrical cables used in low-voltage systems. Practical techniques for determining which cables are prone to premature aging and for evaluating their condition are described. The course contains 11 modules. The first 9 modules describe the body of the cable aging management process. The last two modules, "Cable Failure Assessment" and "Performing a Walkdown o...

2003-11-24T23:59:59.000Z

31

Aging management guideline for commercial nuclear power plants-pumps  

SciTech Connect

This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant pumps important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

Booker, S.; Katz, D.; Daavettila, N.; Lehnert, D. [MDC-Ogden Environmental and Energy Services, Southfield, MI (United States)

1994-03-01T23:59:59.000Z

32

DEGRADATION SUSCEPTIBILITY METRICS AS THE BASES FOR BAYESIAN RELIABILITY MODELS OF AGING PASSIVE COMPONENTS AND LONG-TERM REACTOR RISK  

Science Conference Proceedings (OSTI)

Conventional probabilistic risk assessments (PRAs) are not well-suited to addressing long-term reactor operations. Since passive structures, systems and components are among those for which refurbishment or replacement can be least practical, they might be expected to contribute increasingly to risk in an aging plant. Yet, passives receive limited treatment in PRAs. Furthermore, PRAs produce only snapshots of risk based on the assumption of time-independent component failure rates. This assumption is unlikely to be valid in aging systems. The treatment of aging passive components in PRA does present challenges. First, service data required to quantify component reliability models are sparse, and this problem is exacerbated by the greater data demands of age-dependent reliability models. A compounding factor is that there can be numerous potential degradation mechanisms associated with the materials, design, and operating environment of a given component. This deepens the data problem since the risk-informed management of materials degradation and component aging will demand an understanding of the long-term risk significance of individual degradation mechanisms. In this paper we describe a Bayesian methodology that integrates the metrics of materials degradation susceptibility being developed under the Nuclear Regulatory Commission's Proactive Management of Materials of Degradation Program with available plant service data to estimate age-dependent passive component reliabilities. Integration of these models into conventional PRA will provide a basis for materials degradation management informed by the predicted long-term operational risk.

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

2011-07-17T23:59:59.000Z

33

Prognostics Health Management for Advanced Small Modular Reactor Passive Components  

SciTech Connect

In the United States, sustainable nuclear power to promote energy security is a key national energy priority. Advanced small modular reactors (AdvSMR), which are based on modularization of advanced reactor concepts using non-light-water reactor (LWR) coolants such as liquid metal, helium, or liquid salt may provide a longer-term alternative to more conventional LWR-based concepts. The economics of AdvSMRs will be impacted by the reduced economy-of-scale savings when compared to traditional LWRs and the controllable day-to-day costs of AdvSMRs are expected to be dominated by operations and maintenance costs. Therefore, achieving the full benefits of AdvSMR deployment requires a new paradigm for plant design and management. In this context, prognostic health management of passive components in AdvSMRs can play a key role in enabling the economic deployment of AdvSMRs. In this paper, the background of AdvSMRs is discussed from which requirements for PHM systems are derived. The particle filter technique is proposed as a prognostics framework for AdvSMR passive components and the suitability of the particle filter technique is illustrated by using it to forecast thermal creep degradation using a physics-of-failure model and based on a combination of types of measurements conceived for passive AdvSMR components.

Meyer, Ryan M.; Ramuhalli, Pradeep; Coble, Jamie B.; Mitchell, Mark R.; Wootan, David W.; Hirt, Evelyn H.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

2013-10-18T23:59:59.000Z

34

DOE - Office of Legacy Management -- Elk River Reactor - MN 01  

Office of Legacy Management (LM)

Elk River Reactor - MN 01 FUSRAP Considered Sites Site: Elk River Reactor (MN.01 ) Eliminated from consideration under FUSRAP - Reactor was dismantled and decommissioned by 1974...

35

Calvert Cliffs License Renewal Aging Management Review Reports  

Science Conference Proceedings (OSTI)

This report and associated CD-ROM documents the Calvert Cliffs License Renewal Aging Management Review Reports as of December 1998. The option to operate a nuclear power plant beyond its initial license term of 40 years is an important factor in financial decisions and long-term planning for utility asset management and capital improvement. This report provides other utilities with details of the technical evaluations to support their own preparation of a License Renewal Application (LRA).

1999-06-23T23:59:59.000Z

36

Understanding and managing the effects of battery charger and inverter aging  

SciTech Connect

An aging assessment of battery chargers and inverters was conducted under the auspices of the NRC's Nuclear Plant Aging Research (NPAR) Program. The intentions of this program are to resolve issues related to the aging and service wear of equipment and systems at operating reactor facilities and to assess their impact on safety. Inverters and battery chargers are used in nuclear power plants to perform significant functions related to plant safety and availability. The specific impact of a battery charger or inverter failure varies with plant configuration. Operating experience data have demonstrated that reactor trips, safety injection system actuations, and inoperable emergency core cooling systems have resulted from inverter failures; and dc bus degradation leading to diesel generator inoperability or loss of control room annunication and indication have resulted from battery and battery charger failures. For the battery charger and inverter, the aging and service wear of subcomponents have contributed significantly to equipment failures. This paper summarizes the data and then describes methods that can be used to detect battery charger and inverter degradation prior to failure, as well as methods to minimize the failure effects. In both cases, the managing of battery charger and inverter aging is emphasized. 5 refs.

Gunther, W. (Brookhaven National Lab., Upton, NY (United States)); Aggarwal, S. (Nuclear Regulatory Commission, Washington, DC (United States))

1992-01-01T23:59:59.000Z

37

Understanding and managing the effects of battery charger and inverter aging  

SciTech Connect

An aging assessment of battery chargers and inverters was conducted under the auspices of the NRC`s Nuclear Plant Aging Research (NPAR) Program. The intentions of this program are to resolve issues related to the aging and service wear of equipment and systems at operating reactor facilities and to assess their impact on safety. Inverters and battery chargers are used in nuclear power plants to perform significant functions related to plant safety and availability. The specific impact of a battery charger or inverter failure varies with plant configuration. Operating experience data have demonstrated that reactor trips, safety injection system actuations, and inoperable emergency core cooling systems have resulted from inverter failures; and dc bus degradation leading to diesel generator inoperability or loss of control room annunication and indication have resulted from battery and battery charger failures. For the battery charger and inverter, the aging and service wear of subcomponents have contributed significantly to equipment failures. This paper summarizes the data and then describes methods that can be used to detect battery charger and inverter degradation prior to failure, as well as methods to minimize the failure effects. In both cases, the managing of battery charger and inverter aging is emphasized. 5 refs.

Gunther, W. [Brookhaven National Lab., Upton, NY (United States); Aggarwal, S. [Nuclear Regulatory Commission, Washington, DC (United States)

1992-06-01T23:59:59.000Z

38

RELIABILITY MODELS OF AGING PASSIVE COMPONENTS INFORMED BY MATERIALS DEGRADATION METRICS TO SUPPORT LONG-TERM REACTOR OPERATIONS  

Science Conference Proceedings (OSTI)

Paper describes a methodology for the synthesis of nuclear power plant service data with expert-elicited materials degradation information to estimate the future failure rates of passive components. This method should be an important resource to long-term plant operations and reactor life extension. Conventional probabilistic risk assessments (PRAs) are not well suited to addressing long-term reactor operations. Since passive structures and components are among those for which replacement can be least practical, they might be expected to contribute increasingly to risk in an aging plant; yet, passives receive limited treatment in PRAs. Furthermore, PRAs produce only snapshots of risk based on the assumption of time-independent component failure rates. This assumption is unlikely to be valid in aging systems. The treatment of aging passive components in PRA presents challenges. Service data to quantify component reliability models are sparse, and this is exacerbated by the greater data demands of age-dependent reliability models. Another factor is that there can be numerous potential degradation mechanisms associated with the materials and operating environment of a given component. This deepens the data problem since risk-informed management of component aging will demand an understanding of the long-term risk significance of individual degradation mechanisms. In this paper we describe a Bayesian methodology that integrates metrics of materials degradation susceptibility with available plant service data to estimate age-dependent passive component reliabilities. Integration of these models into conventional PRA will provide a basis for materials degradation management informed by predicted long-term operational risk.

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

2012-05-01T23:59:59.000Z

39

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

40

EPRI Materials Management Matrix Project: Advanced Light-Water Reactor - Pressurized Water Reactor Degradation Matrix - Revision 1  

Science Conference Proceedings (OSTI)

The Advanced Light Water Reactor - Pressurized Water Reactor Degradation Matrix (ALWR PWR DM) is an integral piece of the Electric Power Research Institutes (EPRIs) Materials Management Matrix (MMM) initiative for ALWR designs. The MMM provides a tool to assist the industry in proactive identification and consideration of materials issues and mitigation/management opportunities from the design phase through component fabrication and plant construction to operations and maintenance.

2010-09-22T23:59:59.000Z

Note: This page contains sample records for the topic "reactor aging management" 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

Tritium production analysis and management strategies for a Fluoride-salt-cooled high-temperature test reactor (FHTR)  

E-Print Network (OSTI)

The Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a test reactor concept that aims to demonstrate the neutronics, thermal-hydraulics, materials, tritium management, and to address other reactor operational ...

Rodriguez, Judy N

2013-01-01T23:59:59.000Z

42

Reactor Vessel Head Disposal Campaign for Nuclear Management Company  

SciTech Connect

After establishing a goal to replace as many reactor vessel heads as possible - in the shortest time and at the lowest cost as possible - Nuclear Management Company (NMC) initiated an ambitious program to replace the heads on all six of its pressurized water reactors. Currently, four heads have been replaced; and four old heads have been disposed of. In 2002, NMC began fabricating the first of its replacement reactor vessel heads for the Kewaunee Nuclear Plant. During its fall 2004 refueling outage, Kewaunee's head was replaced and the old head was prepared for disposal. Kewaunee's disposal project included: - Down-ending, - Draining, - Decontamination, - Packaging, - Removal from containment, - On-Site handling, - Temporary storage, - Transportation, - Disposal. The next two replacements took place in the spring of 2005. Point Beach Nuclear Plant (PBNP) Unit 2 and Prairie Island Nuclear Generating Plant (PINGP) Unit 2 completed their head replacements during their scheduled refueling outages. Since these two outages were scheduled so close to each other, their removal and disposal posed some unique challenges. In addition, changes to the handling and disposal programs were made as a result of lessons learned from Kewaunee. A fourth head replacement took place during PBNP Unit 1's refueling outage during the fall of 2005. A number of additional changes took place. All of these changes and challenges are discussed in the paper. NMC's future schedule includes PINGP Unit 1's installation in Spring 2006 and Palisades' installation during 2007. NMC plans to dispose of these two remaining heads in a similar manner. This paper presents a summary of these activities, plus a discussion of lessons learned. (authors)

Hoelscher, H.L.; Closs, J.W. [Nuclear Management Company, LLC, 700 First Street, Hudson, WI 54016 (United States); Johnson, S.A. [Duratek, Inc., 140 Stoneridge Drive, Columbia, SC 29210 (United States)

2006-07-01T23:59:59.000Z

43

COMPONENT DEGRADATION SUSCEPTIBILITIES AS THE BASES FOR MODELING REACTOR AGING RISK  

Science Conference Proceedings (OSTI)

The extension of nuclear power plant operating licenses beyond 60 years in the United States will be necessary if we are to meet national energy needs while addressing the issues of carbon and climate. Characterizing the operating risks associated with aging reactors is problematic because the principal tool for risk-informed decision-making, Probabilistic Risk Assessment (PRA), is not ideally-suited to addressing aging systems. The components most likely to drive risk in an aging reactor - the passives - receive limited treatment in PRA, and furthermore, standard PRA methods are based on the assumption of stationary failure rates: a condition unlikely to be met in an aging system. A critical barrier to modeling passives aging on the wide scale required for a PRA is that there is seldom sufficient field data to populate parametric failure models, and nor is there the availability of practical physics models to predict out-year component reliability. The methodology described here circumvents some of these data and modeling needs by using materials degradation metrics, integrated with conventional PRA models, to produce risk importance measures for specific aging mechanisms and component types. We suggest that these measures have multiple applications, from the risk-screening of components to the prioritization of materials research.

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

2010-07-18T23:59:59.000Z

44

Accident management for indian pressurized heavy water reactors  

Science Conference Proceedings (OSTI)

Indian nuclear power program as of now is mainly based on Pressurized Heavy Water Reactors (PHWRs). Operating Procedures for normal power operation and Emergency Operating Procedures for operational transients and accidents within design basis exist for all Indian PHWRs. In addition, on-site and off-site emergency response procedures are also available for these NPPs. The guidelines needed for severe accidents mitigation are now formally being documented for Indian PHWRs. Also, in line with International trend of having symptom based emergency handling, the work is in advanced stage for preparation of symptom-based emergency operating procedures. Following a plant upset condition; a number of alarms distributed in different information systems appear in the control room to aid operator to identify the nature of the event. After identifying the event, appropriate intervention in the form of event based emergency operating procedure is put into use by the operating staff. However, if the initiating event cannot be unambiguously identified or after the initial event some other failures take place, then the selected event based emergency operating procedure will not be optimal. In such a case, reactor safety is ensured by monitoring safety functions (depicted by selected plant parameters grouped together) throughout the event handling so that the barriers to radioactivity release namely, fuel and fuel cladding, primary heat transport system integrity and containment remain intact. Simultaneous monitoring of all these safety functions is proposed through status trees and this concept will be implemented through a computer-based system. For beyond design basis accidents, event sequences are identified which may lead to severe core damage. As part of this project, severe accident mitigation guidelines are being finalized for the selected event sequences. The paper brings out the details of work being carried out for Indian PHWRs for symptom based event handling and severe accident management. (authors)

Hajela, S.; Grover, R.; Ghadge, S.G.; Bajaj, S.S. [Directorate of Safety, Nuclear Power Corporation of India Limited Nabhikiya Urja Bhawan, Anushakti Nagar, Mumbai-400 094 (India)

2006-07-01T23:59:59.000Z

45

Water inventory management in condenser pool of boiling water reactor  

DOE Patents (OSTI)

An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

Gluntz, Douglas M. (San Jose, CA)

1996-01-01T23:59:59.000Z

46

Water inventory management in condenser pool of boiling water reactor  

DOE Patents (OSTI)

An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

Gluntz, D.M.

1996-03-12T23:59:59.000Z

47

Analytical modeling of core hydraulics and flow management in breeder reactors  

SciTech Connect

An analytical model representing the hydraulic behavior of the primary system of fast breeder nuclear reactors is discussed. A computer code capable of detailing the core flow distribution and characterizing the flow and pressure drop in each reactor component is presented. Application of this method to the reactor core thermal-hydraulic design has allowed optimization of the flow management with consequent upgrading in performance, reduction of unnecessary conservatism and very substantial cost savings. Typical quantitative examples are presented.

Carelli, M.D.; Willis, J.M.

1979-01-01T23:59:59.000Z

48

The Management of Aging in Nuclear Power Plant Concrete Structures  

Science Conference Proceedings (OSTI)

An example of application of structural reliability theory to investigate the impact of ... Cladding and Duct Materials for Advanced Nuclear Recycle Reactors

49

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

50

DOE - Office of Legacy Management -- Westinghouse Advanced Reactors Div  

Office of Legacy Management (LM)

Advanced Reactors Div Advanced Reactors Div Plutonium and Advanced Fuel Labs - PA 10 FUSRAP Considered Sites Site: WESTINGHOUSE ADVANCED REACTORS DIV., PLUTONIUM FUEL LABORATORIES, AND THE ADVANCED FUEL LAB (PA.10 ) Eliminated from further consideration under FUSRAP Designated Name: Not Designated Alternate Name: None Location: Cheswick , Pennsylvania PA.10-1 Evaluation Year: Circa 1987 PA.10-1 PA.10-4 Site Operations: 1960s and 1970s - Produced light water and fast breeder reactor fuels on a development and pilot plant scale. Closed in 1979. PA.10-2 PA.10-3 Site Disposition: Eliminated - Decommissioned and decontaminated under another Federal program. Release condition confirmed by radiological surveys. PA.10-1 PA.10-2 PA.10-3 PA.10-4 PA.10-5 Radioactive Materials Handled: Yes

51

DOE - Office of Legacy Management -- Westinghouse Advanced Reactors...  

Office of Legacy Management (LM)

PA.10-4 Site Operations: 1960s and 1970s - Produced light water and fast breeder reactor fuels on a development and pilot plant scale. Closed in 1979. PA.10-2 PA.10-3 Site...

52

High Flux Isotopes Reactor (HFIR) Cooling Towers Demolition Waste Management  

SciTech Connect

This paper describes the results of a joint initiative between Oak Ridge National Laboratory, operated by UT-Battelle, and Bechtel Jacobs Company, LLC (BJC) to characterize, package, transport, treat, and dispose of demolition waste from the High Flux Isotope Reactor (HFIR), Cooling Tower. The demolition and removal of waste from the site was the first critical step in the planned HFIR beryllium reflector replacement outage scheduled. The outage was scheduled to last a maximum of six months. Demolition and removal of the waste was critical because a new tower was to be constructed over the old concrete water basin. A detailed sampling and analysis plan was developed to characterize the hazardous and radiological constituents of the components of the Cooling Tower. Analyses were performed for Resource Conservation and Recovery Act (RCRA) heavy metals and semi-volatile constituents as defined by 40 CFR 261 and radiological parameters including gross alpha, gross beta, gross gamma, alpha-emitting isotopes and beta-emitting isotopes. Analysis of metals and semi-volatile constituents indicated no exceedances of regulatory limits. Analysis of radionuclides identified uranium and thorium and associated daughters. In addition 60Co, 99Tc, 226Rm, and 228Rm were identified. Most of the tower materials were determined to be low level radioactive waste. A small quantity was determined not to be radioactive, or could be decontaminated. The tower was dismantled October 2000 to January 2001 using a detailed step-by-step process to aid waste segregation and container loading. The volume of waste as packaged for treatment was approximately 1982 cubic meters (70,000 cubic feet). This volume was comprised of plastic ({approx}47%), wood ({approx}38%) and asbestos transite ({approx}14%). The remaining {approx}1% consisted of the fire protection piping (contaminated with lead-based paint) and incidental metal from conduit, nails and braces/supports, and sludge from the basin. The waste, except for the asbestos, was volume reduced via a private contract mechanism established by BJC. After volume reduction, the waste was packaged for rail shipment. This large waste management project successfully met cost and schedule goals.

Pudelek, R. E.; Gilbert, W. C.

2002-02-26T23:59:59.000Z

53

Mechanical properties of thermally aged cast stainless steels from Shippingport reactor components  

SciTech Connect

Thermal embrittlement of static-cast CF-8 stainless steel components from the decommissioned Shippingport reactor has been characterized. Cast stainless steel materials were obtained from four cold-leg check valves, three hot-leg main shutoff valves, and two pump volutes. The actual time-at-temperature for the materials was {approximately}13 y at {approximately}281 C (538 F) for the hot-leg components and {approximately}264 C (507 F) for the cold-leg components. Baseline mechanical properties for as-cast material were determined from tests on either recovery-annealed material, i.e., annealed for 1 h at 550 C and then water quenched, or material from the cooler region of the component. The Shippingport materials show modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength because of relatively low service temperatures and ferrite content of the steel. The procedure and correlations developed at Argonne National Laboratory for estimating mechanical properties of cast stainless steels predict accurate or slightly lower values for Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and J{sub IC} of the materials. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy achieved after long-term aging, were established from materials that were aged further in the laboratory. The results were consistent with the estimates. The correlations successfully predicted the mechanical properties of the Ringhals 2 reactor hot and crossover-leg elbows (CF-8M steel) after service of {approximately} 15 y and the KRB reactor pump cover plate (CF-8) after {approximately} 8 y of service.

Chopra, O.K.; Shack, W.J. [Argonne National Lab., IL (United States)

1995-04-01T23:59:59.000Z

54

Materials Reliability Program: Pressurized Water Reactor Issue Management Tables (MRP-205)  

Science Conference Proceedings (OSTI)

This report provides PWR Issue Management Tables (IMTs) that identify, prioritize, and describe R&D gaps related to degradation issues in PWR Reactor Pressure Vessels (RPVs), Reactor Internals, ASME Class 1 Piping Components, Pressurizers, and Steam Generators. An R&D "Gap" is identified whenever there are needs in the areas of degradation mechanism understanding, mitigation techniques, repair/replacement techniques, or inspection & evaluation technologies to provide reasonable assurance that a component...

2006-11-29T23:59:59.000Z

55

BWRVIP-167: BWR Vessel and Internals Project, Boiling Water Reactor Issue Management Tables  

Science Conference Proceedings (OSTI)

Ongoing issues related to degradation of boiling water reactor (BWR) pressure vessels, reactor internals, and American Society of Mechanical Engineers (ASME) Class 1 piping components have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) issues. This BWR Vessel and Internals Project (BWRVIP) report provides BWR Issue Management Tables that identify, rank, and describe R&D gaps.

2007-03-20T23:59:59.000Z

56

BWRVIP-167NP, Rev. 3: Boiling Water Reactor Issue Management Tables  

Science Conference Proceedings (OSTI)

Nuclear utilities continue to face a number of ongoing issues related to degradation of boiling water reactor (BWR) pressure vessels, reactor internals, and American Society of Mechanical Engineers (ASME) Class 1 piping components. These issues have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) gaps and BWR Vessel and Internals Project (BWRVIP) requirements. The BWR Issue Management Tables in the report are living documents that ...

2013-08-23T23:59:59.000Z

57

BWRVIP-167NP, Revision 2: BWR Vessel and Internals Project, Boiling Water Reactor Issue Management Tables  

Science Conference Proceedings (OSTI)

Nuclear utilities face numerous ongoing issues related to degradation of boiling water reactor (BWR) pressure vessels, reactor internals, and American Society of Mechanical Engineers (ASME) Class 1 piping components. These issues have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) issues and BWR Vessel and Internals Project (BWRVIP) requirements. The BWR Issue Management Tables (IMTs) in the report are living documents that summarize the st...

2010-08-24T23:59:59.000Z

58

Materials Reliability Program: Pressurized Water Reactor Issue Management Table, PWR-IMT Consequence of Failure (MRP-156)  

Science Conference Proceedings (OSTI)

The Industry Initiative on the Management of Materials Issues provides a proactive, safety-focused approach to the management of materials degradation. In support of this initiative, EPRI formed the Materials Degradation Assessment/Issue Management Table Ad-Hoc Committee and developed an Issue Management Table (IMT) for reactor coolant system components. This report provides initial input to the IMT to address the consequences of failure for the identified components in the reactor coolant system for ope...

2005-12-16T23:59:59.000Z

59

Managing Aging Effects on Dry Cask Storage Systems for Extended Long Term  

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

Managing Aging Effects on Dry Cask Storage Systems for Extended Managing Aging Effects on Dry Cask Storage Systems for Extended Long Term Storage and Transportation of Used Fuel Rev0 Managing Aging Effects on Dry Cask Storage Systems for Extended Long Term Storage and Transportation of Used Fuel Rev0 The report is intended to help assess and establish the technical basis for extended long-term storage and transportation of used nuclear fuel. It provides: 1) an overview of the ISFSI license renewal process based on 10 CFR 72 and the guidance provided in NUREG-1927; 2) definitions and terms for structures and components in DCSSs, materials, environments, aging effects, and aging mechanisms; 3) TLAAs and AMPs, respectively, that have been developed for managing aging effects on the SSCs important to safety in the dry cask storage system designs; and 4) AMPs and TLAAs for the SSCs

60

A Review of Information for Managing Aging in Nuclear Power Plants  

SciTech Connect

Age related degradation effects in safety related systems of nuclear power plants should be managed to prevent safety margins from eroding below the acceptable limits provided in plant design bases. The Nuclear Plant Aging Research (NPAR) Pro- gram, conducted under the auspices of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, and other related aging management programs are developing technical information on managing aging. The aging management process central to these efforts consists of three key elements: 1) selecting structures, systems, and components (SSCs) in which aging should be controlled; 2) understanding the mechanisms and rates of degradation in these SSCs; and 3) managing degradation through effective inspection, surveillance, condition monitoring, trending, record keeping, mainten- ance, refurbishment, replacement, and adjustments in the operating environment and service conditions. This document concisely reviews and integrates information developed under the NPAR Program and other aging management studies and other available information related to understanding and managing age-related degradation effects and provides specific refer- ences to more comprehensive information on the same subjects.

WC Morgan; JV Livingston

1995-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor aging management" 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

Materials Reliability Program: Pressurized Water Reactor Issue Management Tables—Revision 2 (MRP-205)  

Science Conference Proceedings (OSTI)

Ongoing issues related to the degradation of pressurized water reactor (PWR) nuclear steam supply system (NSSS) components have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) issues and associated Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) and Steam Generator Management Program (SGMP) requirements.

2010-10-27T23:59:59.000Z

62

Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A)  

Science Conference Proceedings (OSTI)

The Materials Reliability Program (MRP) developed inspection and evaluation (I&E) guidelines for managing long-term aging reactor vessel internal components of pressurized water reactors (PWRs) reactor internals. Specifically, the guidelines are applicable to reactor vessel internal structural components; they do not address fuel assemblies, reactivity control assemblies, or welded attachments to the reactor vessel.

2011-12-23T23:59:59.000Z

63

Supplying the nuclear arsenal: Production reactor technology, management, and policy, 1942--1992  

SciTech Connect

This book focuses on the lineage of America`s production reactors, those three at Hanford and their descendants, the reactors behind America`s nuclear weapons. The work will take only occasional sideways glances at the collateral lines of descent, the reactor cousins designed for experimental purposes, ship propulsion, and electric power generation. Over the decades from 1942 through 1992, fourteen American production reactors made enough plutonium to fuel a formidable arsenal of more than twenty thousand weapons. In the last years of that period, planners, nuclear engineers, and managers struggled over designs for the next generation of production reactors. The story of fourteen individual machines and of the planning effort to replace them might appear relatively narrow. Yet these machines lay at the heart of the nation`s nuclear weapons complex. The story of these machines is the story of arming the winning weapon, supplying the nuclear arms race. This book is intended to capture the history of the first fourteen production reactors, and associated design work, in the face of the end of the Cold War.

Carlisle, R.P.; Zenzen, J.M.

1994-01-01T23:59:59.000Z

64

Aging management guideline for commercial nuclear power plants-stationary batteries. Final report  

Science Conference Proceedings (OSTI)

The Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant stationary batteries important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

Berg, R.; Shao, J.; Krencicki, G.; Giachetti, R. [Multiple Dynamics Corp., Southfield, MI (United States)

1994-03-01T23:59:59.000Z

65

Aging management guideline for commercial nuclear power plants - tanks and pools  

Science Conference Proceedings (OSTI)

Continued operation of nuclear power plants for periods that extend beyond their original 40-year license period is a desirable option for many U.S. utilities. U.S. Nuclear Regulatory Commission (NRC) approval of operating license renewals is necessary before continued operation becomes a reality. Effective aging management for plant components is important to reliability and safety, regardless of current plant age or extended life expectations. However, the NRC requires that aging evaluations be performed and the effectiveness of aging management programs be demonstrated for components considered within the scope of license renewal before granting approval for operation beyond 40 years. Both the NRC and the utility want assurance that plant components will be highly reliable during both the current license term and throughout the extended operating period. In addition, effective aging management must be demonstrated to support Maintenance Rule (10 CFR 50.65) activities.

Blocker, E.; Smith, S.; Philpot, L.; Conley, J.

1996-02-01T23:59:59.000Z

66

Aging Management Guidelines for Commercial Nuclear Power Plants -- Electrical and Mechanical Penetrations  

Science Conference Proceedings (OSTI)

As part of the application process for license renewal, nuclear utilities must perform an evaluation to confirm that they have appropriately considered aging effects on plant components within the scope of the License Renewal Rule. This report provides guidelines for managing relevant degradation mechanisms applicable to electrical and mechanical penetrations in nuclear power plants. Use of these guidelines will provide utilities with a basis for verifying that effective methods for managing age-related ...

2002-04-25T23:59:59.000Z

67

Managing Risk and Improving Financial Performance for an Aging Turbo-Generator Fleet  

Science Conference Proceedings (OSTI)

This document is a brief guide for maintenance staff on effectively using the PowerPoint slideshow, A Proposal to Better Manage Risk and Improve Financial Performance of an Aging Turbo-Generator Fleet, as a starting point for discussing the issue with plant management.

2008-11-03T23:59:59.000Z

68

Status of the Integral Fast Reactor fuel cycle demonstration and waste management practices  

SciTech Connect

Over the past few years, Argonne National Laboratory has been preparing for the demonstration of the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety and operations, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle, which will be demonstrated at Argonne-West in Idaho, employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The required facility modifications and process equipment for the demonstration are nearing completion. Their status and the results from initial fuel fabrication work, including the waste management aspects, are presented. Additionally, estimated compositions of the various process waste streams have been made, and characterization and treatment methods are being developed. The status of advanced waste processing equipment being designed and fabricated is described.

Benedict, R.W.; Goff, K.M.; McFarlane, H.F.

1994-07-01T23:59:59.000Z

69

Aging  

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

Aging Name: davida Status: NA Age: NA Location: NA Country: NA Date: Around 1993 Question: What causes aging to occur? Replies: A deep and far-reaching question Aging is...

70

Materials Reliability Program: Pressurized Water Reactor Issue Management Tables - Revision 3 (MRP-205)  

Science Conference Proceedings (OSTI)

Nuclear utilities continue to face a number of ongoing issues related to degradation of pressurized water reactor (PWR) nuclear steam supply system (NSSS) components. These issues have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) issues and associated EPRI Materials Reliability Program (MRP) and Steam Generator Management Program (SGMP) requirements.BackgroundA comprehensive, integrated ...

2013-12-03T23:59:59.000Z

71

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

72

EPRI Materials Management Matrix Project: U.S.--Advanced Pressurized Water Reactor Materials Management Tables  

Science Conference Proceedings (OSTI)

The Electric Power Research Institutes (EPRIs) Advanced Nuclear Technology (ANT) Program has initiated a Materials Management Matrix (MMM) initiative to systematically assess new plant designs and identify gaps and opportunities that, if addressed at appropriate times in the life cycle of a plant, could significantly improve performance of the materials used in new plant designs. Products developed by this initiative are intended to be living reports that will be periodically updated through the licensin...

2011-06-17T23:59:59.000Z

73

Advanced Test Reactor Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables  

SciTech Connect

U.S. Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Advanced Test Reactor Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. U.S. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool to develop the radioactive waste management basis.

Lisa Harvego; Brion Bennett

2011-11-01T23:59:59.000Z

74

A Project Management and Systems Engineering Structure for a Generation IV Very High Temperature Reactor  

DOE Green Energy (OSTI)

The Very High Temperature Reactor (VHTR) will be an advanced, very high temperature (approximately 1000o C. coolant outlet temperature), gas cooled nuclear reactor and is the nearest term of six Generation IV reactor technologies for nuclear assisted hydrogen production. In 2001, the Generation IV International Forum (GIF), a ten nation international forum working together with the Department of Energy’s (DOE) Nuclear Energy Research Advisory Committee (NERAC), agreed to proceed with the development of a technology roadmap and identified the next generation of nuclear reactor systems for producing new sources of power. Since a new reactor has not been licensed in the United States since the 1970s, the risks are too large for a single utility to assume in the development of an unprecedented Generation IV reactor. The government must sponsor and invest in the research to resolve major first of a kind (FOAK) issues through a full-scale demonstration prior to industry implementation. DOE’s primary mission for the VHTR is to demonstrate nuclear reactor assisted cogeneration of electricity and hydrogen while meeting the Generation IV goals for safety, sustainability, proliferation resistance and physical security and economics. The successful deployment of the VHTR as a demonstration project will aid in restarting the now atrophied U.S. nuclear power industry infrastructure. It is envisioned that VHTR project participants will include DOE Laboratories, industry partners such as designers, constructors, manufacturers, utilities, and Generation IV international countries. To effectively mange R&D, engineering, procurement, construction, and operation for this multi-organizational and technologically complex project, systems engineering will be used extensively to ensure delivery of the final product. Although the VHTR is an unprecedented FOAK system, the R&D, when assessed using the Office of Science and Technology Gate Model, falls primarily in the 3rd - Exploratory Development, 4th – Advanced Development, and 5th- Engineering Development stages of maturity rather than in the basic and viability stages. Therefore the R&D must be controlled and project driven from the top down to address specific issues of feasibility, proof of design or support of engineering. The design evolution must be through the systems approach including an iterative process of high-level requirements definition, engineering to focus R&D to verify feasibility, requirements development and conceptual design, R&D to verify design and refine detailed requirements for final detailed design. This paper will define a framework for project management and application of systems engineering at the Idaho National Engineering and Environmental Laboratory (INEEL). The VHTR Project includes an overall reactor design and construction activity and four major supporting activities: fuel development and qualification, materials selection and qualification, NRC licensing and regulatory support, and the hydrogen production plant.

Ed Gorski; Dennis Harrell; Finis Southworth

2004-09-01T23:59:59.000Z

75

Aging Management of Nuclear Power Plant Concrete Structures - Overview and Suggested Research Topics  

SciTech Connect

Nuclear power plant concrete structures are described and their operating experience noted. Primary considerations related to management of their aging are noted and an indication of their status provided: degradation mechanisms, damage models, and material performance; assessment and remediation (i.e., component selection, in-service inspection, nondestructive examinations, and remedial actions); and estimation of performance at present or some future point in time (i.e., application of structural reliability theory to the design and optimization of in-service inspection/maintenance strategies, and determination of the effects of degradation on plant risk). Several activities are identified that provide background information and data on areas of concern with respect to nondestructive examination of nuclear power plant concrete structures: inspection of thick-walled, heavily-reinforced sections, basemats, and inaccessible areas of the containment metallic pressure boundary. Topics are noted where additional research would be of benefit to aging management of nuclear power plant concrete structures.

Naus, Dan J [ORNL

2008-01-01T23:59:59.000Z

76

Prioritization of reactor control components susceptible to fire damage as a consequence of aging  

SciTech Connect

The Fire Vulnerability of Aged Electrical Components Test Program is to identify and assess issues of plant aging that could lead to an increase in nuclear power plant risk because of fires. Historical component data and prior analyses are used to prioritize a list of components with respect to aging and fire vulnerability and the consequences of their failure on plant safety systems. The component list emphasizes safety system control components, but excludes cables, large equipment, and devices encompassed in the Equipment Qualification (EQ) program. The test program selected components identified in a utility survey and developed test and fire conditions necessary to maximize the effectiveness of the test program. Fire damage considerations were limited to purely thermal effects.

Lowry, W.; Vigil, R. [Science and Engineering Associates, Inc., Albuquerque, NM (United States); Nowlen, S. [Sandia National Labs., Albuquerque, NM (United States)

1994-01-01T23:59:59.000Z

77

Feasibility of Burning First- and Second-Generation Plutonium in Pebble Bed High-Temperature Reactors  

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Fuel Management of Reactors Other Than Light Water Reactors

J. B. M. De Haas; J. C. Kuijper

78

Modeling of Late Blooming Phases and Precipitation Kinetics in Aging Reactor Pressure Vessel (RPV) Steels  

Science Conference Proceedings (OSTI)

The principle work at the atomic scale is to develop a predictive quantitative model for the microstructure evolution of RPV steels under thermal aging and neutron radiation. We have developed an AKMC method for the precipitation kinetics in bcc-Fe, with Cu, Ni, Mn and Si being the alloying elements. In addition, we used MD simulations to provide input parameters (if not available in literature). MMC simulations were also carried out to explore the possible segregation/precipitation morphologies at the lattice defects. First we briefly describe each of the simulation algorithms, then will present our results.

Yongfeng Zhang; Pritam Chakraborty; S. Bulent Biner

2013-09-01T23:59:59.000Z

79

Aging Management Guideline for commercial nuclear power plants: Battery chargers, inverters and uninterruptible power supplies. Final report  

SciTech Connect

This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant battery chargers, inverters and uninterruptible power supplies important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already, experienced) and aging management program activities to the more generic results and recommendations presented herein.

Berg, R.; Stroinski, M.; Giachetti, R. [Multiple Dynamics Corp., Southfield, MI (United States)

1994-02-01T23:59:59.000Z

80

Experience Based Interview Process for Power Plant Management: With a Pilot Application to Aging of Outage Support Equipment  

Science Conference Proceedings (OSTI)

Plant managers are responsible for ensuring that their facilities meet strategic asset management goals in an increasingly competitive power generation marketplace. This report describes an experience based interview (EBI) method for capturing the knowledge of plant staff to more efficiently identify technical or operational issues and make management decisions that protect assets. Included is a demonstration of the method in a pilot application to aging of support equipment that could impact outages in ...

1999-04-05T23:59:59.000Z

Note: This page contains sample records for the topic "reactor aging management" 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

Management of transuranics using the Integral Fast Reactor (IFR) fuel cycle  

Science Conference Proceedings (OSTI)

The 50 years of activities following the discovery of self-sustaining fission chains have given rise to a buildup of roughly 900 tons of manmade transuranics. Of the total, about 260 tons of Pu{sup 239} were generated for use in weapons while the remainder were generated as a byproduct of electrical power produced worldwide by the commercial thermal nuclear power industry. What is to be done with these actinides? The options for disposition include interminable storage, burial, or recycle for use. The pros and cons of each option are being vigorously debated regarding the impact upon the issues of human and ecological risk -- both current and future; weapons proliferation potential -- both current and future; and total life cycle benefits and costs. As to the options for utilization, commercial uses for actinides (uranium and transuranics) are of limited diversity. The actinides have in the past and will in the future find application in large scale mostly by virtue of their ability to release energy through fission, and here their utility is unmatched -- whether the application be in commercial electricity generation or in armaments. The integral Fast Reactor (IFR) fuel cycle offers a number of features for management of the current and future burden of manmade transuranic materials and for capturing the energy content of the U{sup 238}. These features are discussed here.

Wade, D.C.

1994-01-01T23:59:59.000Z

82

The use of PRA (Probabilistic Risk Assessment) in the management of safety issues at the High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

The High Flux Isotope reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988, a level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 {times} 10{sup {minus}4}. In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic, wind, and fires. The overall external event contribution to core damage frequency is about 138% of the internal event initiated contribution and is dominated by wind initiators. The PRA has provided a basis for the management of a wide range of safety and operation issues at the HFIR. 3 refs., 4 figs., 2 tabs.

Flanagan, G.F.

1990-01-01T23:59:59.000Z

83

The (safety-related) heat exchangers aging management guideline for commercial nuclear power plants, and developments since 1994  

Science Conference Proceedings (OSTI)

The US Department of Energy (DOE), in cooperation with the Electric Power Research Institute (EPRI) and US nuclear power plant utilities, is preparing a series of aging management guidelines (AMGs) for commodity types of components (e.g., heat exchangers, electrical cable and terminations, pumps). Commodities are included in this series based on their importance to continued nuclear plant operation and license renewal. The AMGs contain a detailed summary of operating history, stressors, aging mechanisms, and various types of maintenance and surveillance practices that can be combined to create an effective aging management program. Each AMG is intended for use by the systems engineers and plant maintenance staff (i.e., an AMG is intended to be a hands-on technical document rather than a licensing document). The heat exchangers AMG, published in June 1994, includes the following information of interest to nondestructive examination (NDE) personnel: aging mechanisms determined to be non-significant for all applications; aging mechanisms determined to be significant for some applications; effective conventional programs for managing aging; and effective unconventional programs for managing aging. Since the AMG on heat exchangers was published four years ago, a brief review has been conducted to identify emerging regulatory issues, if any. The results of this review and lessons learned from the collective set of AMGs are presented.

Clauss, J.M.

1998-08-01T23:59:59.000Z

84

Program management plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee  

SciTech Connect

The primary mission of the Molten Salt Reactor Experiment (MSRE) Remediation Project is to effectively implement the risk-reduction strategies and technical plans to stabilize and prevent further migration of uranium within the MSRE facility, remove the uranium and fuel salts from the system, and dispose of the fuel and flush salts by storage in appropriate depositories to bring the facility to a surveillance and maintenance condition before decontamination and decommissioning. This Project Management Plan (PMP) for the MSRE Remediation Project details project purpose; technical objectives, milestones, and cost objectives; work plan; work breakdown structure (WBS); schedule; management organization and responsibilities; project management performance measurement planning, and control; conduct of operations; configuration management; environmental, safety, and health compliance; quality assurance; operational readiness reviews; and training.

NONE

1996-09-01T23:59:59.000Z

85

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

86

DOE - Office of Legacy Management -- Reactor Site - Fort Belvoir - VA 0-02  

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

Reactor Site - Fort Belvoir - VA Reactor Site - Fort Belvoir - VA 0-02 FUSRAP Considered Sites Site: REACTOR SITE - FORT BELVOIR (VA.0-02 ) Eliminated from further consideration under FUSRAP - Referred to DOD Designated Name: Not Designated Alternate Name: None Location: Fort Belvoir , Virginia VA.0-02-1 Evaluation Year: 1987 VA.0-02-1 Site Operations: No evidence of AEC involvement with reactor operations. AEC conducted health and safety inspections of this site. Probably a licensed operation. VA.0-02-1 Site Disposition: Eliminated - Referred to DOD VA.0-02-1 Radioactive Materials Handled: Reactor fuel Primary Radioactive Materials Handled: Reactor Fuel Radiological Survey(s): Health and safety inspections VA.0-02-1 Site Status: Eliminated from further consideration under FUSRAP - Referred to DOD VA.0-02-1

87

Plant Support Engineering: Aging Management Program Development Guidance for Instrument and Control Cable Systems for Nuclear Power Plants  

Science Conference Proceedings (OSTI)

This document provides guidance for developing and implementing a cable aging management program for low-voltage instrument and control cable circuits in nuclear power plants. Guidance is provided for identifying cables located in adverse localized environments and determining if those environments have caused significant cable circuit degradation.

2010-11-29T23:59:59.000Z

88

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

89

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...

90

Influence of long-term thermal aging on the microstructural evolution of nuclear reactor pressure vessel materials: An atom probe study  

Science Conference Proceedings (OSTI)

Atom probe field ion microscopy (APFIM) investigations of the microstructure of unaged (as-fabricated) and long-term thermally aged ({approximately} 100,000 h at 280 C) surveillance materials from commercial reactor pressure vessel steels were performed. This combination of materials and conditions permitted the investigation of potential thermal-aging effects. This microstructural study focused on the quantification of the compositions of the matrix and carbides. The APFIM results indicate that there was no significant microstructural evolution after a long-term thermal exposure in weld, plate, or forging materials. The matrix depletion of copper that was observed in weld materials was consistent with the copper concentration in the matrix after the stress-relief heat treatment. The compositions of cementite carbides aged for 100,000 h were compared with the Thermocalc{trademark} prediction. The APFIM comparisons of materials under these conditions are consistent with the measured change in mechanical properties such as the Charpy transition temperature.

Pareige, P.; Russell, K.F.; Stoller, R.E.; Miller, M.K. [Oak Ridge National Lab., TN (United States)

1998-03-01T23:59:59.000Z

91

DOE - Office of Legacy Management -- West Milton Reactor Site - NY 21  

Office of Legacy Management (LM)

Milton Reactor Site - NY 21 Milton Reactor Site - NY 21 FUSRAP Considered Sites Site: West Milton Reactor Site (NY.21) Designated Name: Alternate Name: Location: Evaluation Year: Site Operations: Site Disposition: Radioactive Materials Handled: Primary Radioactive Materials Handled: Radiological Survey(s): Site Status: This site is one of a group of 77 FUSRAP considered sites for which few, if any records are available in their respective site files to provide an historical account of past operations and their relationship, if any, with MED/AEC operations. Reviews of contact lists, accountable station lists, health and safety records and other documentation of the period do not provide sufficient information to warrant further search of historical records for information on these sites. These site files remain "open" to

92

The Small Modular Liquid Metal Cooled Reactor: A New Approach to Proliferation Risk Management  

DOE Green Energy (OSTI)

There is an ongoing need to supply energy to small markets and remote locations with limited fossil fuel infrastructures. The Small, Modular, Liquid-Metal-Cooled Reactor, also referred to as SSTAR (Small, Secure, Transportable, Autonomous Reactor), can provide reliable and cost-effective electricity, heat, fresh water, and potentially hydrogen transportation fuels for these markets. An evaluation of a variety of reactor designs indicates that SSTAR, with its secure, long-life core, has many advantages for deployment into a variety of national and international markets. In this paper, we describe the SSTAR concept and its approach to safety, security, environmental and non-proliferation. The system would be design-certified using a new license-by-test approach, and demonstrated for commercial deployment anywhere in the world. The project addresses a technology development need (i.e., a small secure modular system for remote sites) that is not otherwise addressed in other currently planned research programs.

Smith, C F; Crawford, D; Cappiello, M; Minato, A; Herczeg, J W

2003-11-12T23:59:59.000Z

93

Plant Support Engineering: Aging Effects for Structures and Structural Components (Structural Tools)  

Science Conference Proceedings (OSTI)

As part of the application process for license renewal (LR), nuclear utilities must perform an evaluation to confirm that they have appropriately considered any aging effects on plant components within the scope of the License Renewal Rule. This report identifies structures and structural components subject to aging management review (AMR) for generic pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear plants. The report also describes aging mechanisms that might impact each component.

2007-12-21T23:59:59.000Z

94

Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175)  

Science Conference Proceedings (OSTI)

This report provides screening criteria and their technical bases for age-related degradation evaluation of Pressurized Water Reactor (PWR) internals component items. It is a key element in an overall strategy that uses knowledge of internals design, materials, and material properties and applies screening methodologies for known age-related degradation mechanisms to manage the effects of aging in PWR internals.

2005-12-12T23:59:59.000Z

95

Guidelines for Managing Reactor Vessel Material Uncertainties: Part 1: General Approach Part 2: Implementation Guide  

Science Conference Proceedings (OSTI)

Uncertainties about reactor vessel material toughness properties can be a concern for utilities when characterizing vessel integrity. In addition, recent emphasis on variability in material chemistry and initial toughness properties has added to regulatory concerns. This two-part guidelines document provides a general approach (Part 1) for dealing with weld metal property variability and material uncertainties and demonstrates examples of different approaches (Part 2) for dealing with these uncertainties...

1997-04-30T23:59:59.000Z

96

A Parametric Study of the DUPIC Fuel Cycle to Reflect Pressurized Water Reactor Fuel Management Strategy  

SciTech Connect

For both pressurized water reactor (PWR) and Canada deuterium uranium (CANDU) tandem analysis, the Direct Use of spent PWR fuel In CANDU reactor (DUPIC) fuel cycle in a CANDU 6 reactor is studied using the DRAGON/DONJON chain of codes with the ENDF/B-V and ENDF/B-VI libraries. The reference feed material is a 17 x 17 French standard 900-MW(electric) PWR fuel. The PWR spent-fuel composition is obtained from two-dimensional DRAGON assembly transport and depletion calculations. After a number of years of cooling, this defines the initial fuel nuclide field in the CANDU unit cell calculations in DRAGON, where it is further depleted with the same neutron group structure. The resulting macroscopic cross sections are condensed and tabulated to be used in a full-core model of a CANDU 6 reactor to find an optimized channel fueling rate distribution on a time-average basis. Assuming equilibrium refueling conditions and a particular refueling sequence, instantaneous full-core diffusion calculations are finally performed with the DONJON code, from which both the channel power peaking factors and local parameter effects are estimated. A generic study of the DUPIC fuel cycle is carried out using the linear reactivity model for initial enrichments ranging from 3.2 to 4.5 wt% in a PWR. Because of the uneven power histories of the spent PWR assemblies, the spent PWR fuel composition is expected to differ from one assembly to the next. Uneven mixing of the powder during DUPIC fuel fabrication may lead to uncertainties in the composition of the fuel bundle and larger peaking factors in CANDU. A mixing method for reducing composition uncertainties is discussed.

Rozon, Daniel; Shen Wei [Institut de Genie Nucleaire (Canada)

2001-05-15T23:59:59.000Z

97

PWR Primary-Side Gas Management in Advanced Light Water Reactors  

Science Conference Proceedings (OSTI)

The designs for advanced light water reactors (ALWRs) have incorporated new water chemistry controls that have been developed over the past few decades to improve material and equipment reliability and fuel performance and to minimize radionuclide production and transport. It is important to ensure that the new designs operate within ranges that are considered safe based on current knowledge and that industry guidance for normal operation, startup, and shutdown are updated to account for expanding ...

2013-07-17T23:59:59.000Z

98

Advanced Core Design And Fuel Management For Pebble-Bed Reactors  

Science Conference Proceedings (OSTI)

A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

2004-10-01T23:59:59.000Z

99

Plant Engineering: Cable Aging Management Training: Identification of Adverse Environment, and Introduction to Visual/ Tactile Asses sment of Cable  

Science Conference Proceedings (OSTI)

Two cable aging management training modules are included that introduce the viewer to adverse environments and their effects on cable polymers and to visual tactile assessment of cables to determine if they have degraded from the adverse environments. The types of environments that are adverse with respect to cable materials and the locations in a plant where they might exist are provided. The visual/tactile presentation shows how basic properties and visual attributes may be used to determine ifcables a...

2011-10-19T23:59:59.000Z

100

EPRI Materials Management Matrix Project: KHNP Advanced Pressurized Water Reactor (APR1400) Materials Management Tables -- Revision 0  

Science Conference Proceedings (OSTI)

Experience gained through years of operating nuclear plants has shown that materials performance issues can be a significant concern related to long-term plant operations. While a number of concerns remain, industry efforts to address materials performance issues at operating plants has led to important advances in both the underlying scientific understanding of materials degradation and the implementation of practical mitigation and management technologies.

2011-11-29T23:59:59.000Z

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


101

The effect of thermal aging and boiling water reactor environment on Type 316L stainless steel welds  

E-Print Network (OSTI)

The thermal aging and consequent embrittlement of materials are ongoing issues in cast stainless steels and duplex stainless steels. Spinodal decomposition is largely responsible for the well known "475°C" embrittlement ...

Lucas, Timothy R

2011-01-01T23:59:59.000Z

102

Use of Thorium in Light Water Reactors  

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Use of Alternate Fuels in Light Water Reactors

Michael Todosow; A. Galperin; S. Herring; M. Kazimi; T. Downar; A. Morozov

103

Light Water Reactor Sustainability (LWRS) Program – Non-Destructive Evaluation (NDE) R&D Roadmap for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants  

SciTech Connect

The purpose of the non-destructive evaluation (NDE) R&D Roadmap for Cables is to support the Materials Aging and Degradation (MAaD) R&D pathway. The focus of the workshop was to identify the technical gaps in detecting aging cables and predicting their remaining life expectancy. The workshop was held in Knoxville, Tennessee, on July 30, 2012, at Analysis and Measurement Services Corporation (AMS) headquarters. The workshop was attended by 30 experts in materials, electrical engineering, U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy (DOE) National Laboratories (Oak Ridge National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and Idaho National Engineering Laboratory), NDE instrumentation development, universities, commercial NDE services and cable manufacturers, and Electric Power Research Institute (EPRI). The motivation for the R&D roadmap comes from the need to address the aging management of in-containment cables at nuclear power plants (NPPs).

Simmons, Kevin L.; Ramuhalli, Pradeep; Brenchley, David L.; Coble, Jamie B.; Hashemian, Hash; Konnik, Robert; Ray, Sheila

2012-09-14T23:59:59.000Z

104

Plant Support Engineering: Aging Management Program Guidance for Medium- Voltage Cable Systems for Nuclear Power Plants  

Science Conference Proceedings (OSTI)

Regulatory and management concern regarding the reliability of medium-voltage cable systems at nuclear plants has been increasing for 5–10 years. The staff of the U.S. Nuclear Regulatory Commission (NRC) are concerned that wetted (up to and including submergence) medium-voltage cable circuits may be degrading to the point at which multiple cable circuits may fail when called on to perform functions affecting safety. Utility managers are concerned that cables may fail, causing adverse safety consequences ...

2010-06-02T23:59:59.000Z

105

Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning  

Science Conference Proceedings (OSTI)

This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage.

Shah, V.N.; Ware, A.G.; Porter, A.M.

1997-03-01T23:59:59.000Z

106

Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 6, Decontamination and decommissioning, accident management, TMI-2  

SciTech Connect

This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. 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. This report, Volume 6, discusses decontamination and decommissioning, accident management, and the Three Mile Island-2 reactor accident. Thirteen reports have been cataloged separately.

Weiss, A. J. [comp.

1988-02-01T23:59:59.000Z

107

AECL/U.S. INERI - Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Power Reactors Fuel Requirements and Down-Select Report  

SciTech Connect

This report documents the first milestone of the International Nuclear Energy Research Initiative (INERI) U.S./Euratom Joint Proposal 1.8 entitled “Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Light-Water Reactors.” The milestone represents the assessment and preliminary study of a variety of fuels that hold promise as transmutation and minor actinide burning fuel compositions for light-water reactors. The most promising fuels of interest to the participants on this INERI program have been selected for further study. These fuel compositions are discussed in this report.

William Carmack; Randy Fielding; Pavel Medvedev; Mitch Meyer

2005-08-01T23:59:59.000Z

108

How gender, age, and geography influence the utilization of radiation therapy in the management of malignant melanoma  

SciTech Connect

Purpose: Comparing radiation therapy utilization rates (RTUR) to those predicted by best evidence is a useful measure of the equity and accessibility of service delivery. In this study the RTUR for melanoma was established for British Columbia, Canada, and compared with the rate suggested by the evidence. Demographic variables, specifically age, gender, and geography that influenced the RTUR were examined with a view to identifying methods of improving underutilization. Methods and Materials: The RTUR in the management of malignant melanoma was taken from British Columbia Cancer registry data for 1986 to 1998. Variations in utilization based on age, gender, health authority, stage of disease, and referral patterns were analyzed. Results: An RTUR of 11% was identified. This was consistent over time. Referral rates decreased between 1986 and 1998. RT is used mostly for later stage disease. Males were more likely to receive RT than females, related to later stage of disease in men. Referral rates decreased, but RTUR for referred cases increased, in health authorities that did not have a cancer center. Conclusions: Use of RT is influenced by age and by stage of disease. Overall RTUR in British Columbia is lower than suggested by best evidence. Referral patterns are influenced by geography. RTUR was higher in males, consistent with a different pattern of disease in males compared with females.

French, John [Radiation Therapy, British Columbia Cancer Agency, Vancouver Centre, Vancouver, British Columbia (Canada)]. E-mail: jfrench@bccancer.bc.ca; McGahan, Colleen [Population and Preventive Oncology, British Columbia Cancer Agency, Vancouver Centre, Vancouver, British Columbia (Canada); Duncan, Graeme [Department of Radiation Oncology, British Columbia Cancer Agency, Vancouver Centre, Vancouver, British Columbia (Canada); Lengoc, Sonca [Radiation Therapy, British Columbia Cancer Agency, Vancouver Centre, Vancouver, British Columbia (Canada); Soo, Jenny [Radiation Therapy, British Columbia Cancer Agency, Vancouver Centre, Vancouver, British Columbia (Canada); Cannon, Jerry [Radiation Therapy, British Columbia Cancer Agency, Vancouver Centre, Vancouver, British Columbia (Canada)

2006-11-15T23:59:59.000Z

109

A high converter concept for fuel management with blanket fuel assemblies in boiling water reactors  

SciTech Connect

Studies on the natural Uranium saving and waste reduction potential of a multiple-plant BWR system were performed. The BWR High Converter system should enable a multiple recycling of MOX fuel in current BWR plants by introducing blanket fuel assemblies and burning Uranium and MOX fuel separately. The feasibility of Uranium cores with blankets and full-MOX cores with Plutonium qualities as low as 40% were studied. The power concentration due to blanket insertion is manageable with modern fuel and acceptable values for the thermal limits and reactivity coefficients were obtained. While challenges remain, full-MOX cores also complied with the main design criteria. The combination of Uranium and Plutonium burners in appropriate proportions could enable obtaining as much as 40% more energy out of Uranium ore. Moreover, a proper adjustment of blanket average stay and Plutonium qualities could lead to a system with nearly no Plutonium left for final disposal. The achievement of such goals with current light water technology makes the BWR HC concept an attractive option to improve the fuel cycle until Gen-IV designs are mature. (authors)

Martinez-Frances, N.; Timm, W.; Rossbach, D. [AREVA, AREVA NP, Erlangen (Germany)

2012-07-01T23:59:59.000Z

110

Stress-assisted, microbial-induced corrosion of stainless steel primary piping and other aging issues at the Omega West Reactor  

SciTech Connect

After the discovery of cooling system leak of about 284 liters per twenty-four (24) hour period, an investigation determined that the 76.2-cm diameter, 33.5-m long stainless-steel (304) OWR delay line was losing water at the same nominal rate. An excavation effort revealed that a circumferential crack, approximately 0.0025 cm in width, extended around the bottom half of the delay line. In addition, other evidence of what appeared to be microcracking and pitting that originated at random nucleated sites around the pipe were also found. Results of destructive analysis and nondestructive testing allowed Los Alamos staff to conclude that the direct cause for the main crack and other pitting resulted from stress-assisted, microbial-induced corrosion of the stainless steel primary piping. The results also indicated that microbial action from bacteria that are normally present in earth can be extremely harmful to stainless- steel piping under certain conditions. Other potential problems that could have also eventually led to a permanent shutdown of the OWR were discussed. These problems, although never encountered nor associated with the current shutdown, were identified in aging studies and are associated with: (1) the water-cooled, bismuth gamma-ray shield and, (2) the aluminum thermal column head seal that prevents reactor vessel water from entering into the graphite-filled thermal column.

Andrade, A.

1995-07-01T23:59:59.000Z

111

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

112

CRAD, Configuration Management - Oak Ridge National Laboratory...  

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

Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor February...

113

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

114

B Reactor | Department of Energy  

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

Operational Management » History » Manhattan Project » Signature Operational Management » History » Manhattan Project » Signature Facilities » B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first large-scale plutonium production reactor. As at Oak Ridge, the need for labor turned Hanford into an atomic boomtown, with the population reaching 50,000 by summer 1944. Similar to the X-10 Graphite Reactor at Oak Ridge in terms of loading and unloading fuel, the B Reactor was built on a much larger scale and used water rather than air as a coolant. Whereas the X-10 had an initial design output of 1,000 kilowatts, the B Reactor was designed to operate at 250,000 kilowatts. Consisting of a 28- by 36-foot, 1,200-ton graphite cylinder lying on its side, the reactor was penetrated through its

115

Plant Engineering, Aging Management Program Guidance for Medium-Voltage Cable Systems for Nuclear Power Plants, Revision 1  

Science Conference Proceedings (OSTI)

This report was prepared in response to regulatory and management concern regarding the reliability of medium-voltage cable systems at nuclear plants. The staff of the U.S. Nuclear Regulatory Commission (NRC) have been concerned that wetted (up to and including submergence) medium-voltage cable circuits may be degrading to the point at which multiple cable circuits may fail when called on to perform functions affecting safety. A utility manager’s concern is that cables may fail, causing ...

2013-06-03T23:59:59.000Z

116

The effects of aging on BWR core isolation cooling systems  

Science Conference Proceedings (OSTI)

A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling (RCIC) system in commercial Boiling Water Reactors (BWRs). This study is part of the Nuclear Plant Aging Research (NPAR) program sponsored by the US Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. The failure data from national databases, as well as plant specific data were reviewed and analyzed to understand the effects of aging on the RCIC system. This analysis identified important components that should receive the highest priority in terms of aging management. The aging characterization provided information on the effects of aging on component failure frequency, failure modes, and failures causes. Current inspection, surveillance, and monitoring practices were also reviewed.

Lee, B.S. [Brookhaven National Lab., Upton, NY (United States)

1994-10-01T23:59:59.000Z

117

Neutronics and Depletion Methods for Parametric Studies of Fluoride Salt Cooled High Temperature Reactors with Slab Fuel Geometry and Multi-Batch Fuel Management Schemes  

SciTech Connect

The Advanced High Temperature Reactor (AHTR) is a 3400 MWth fluoride salt cooled high temperature reactor (FHR) that uses TRISO particle fuel compacted into slabs rather than spherical fuel pebbles or cylindrical fuel compacts. Simplified methods are required for parametric design studies such that analyzing the entire feasible design space for an AHTR is tractable. These simplifications include fuel homogenization techniques to increase the speed of neutron transport calculations in depletion analysis and equilibrium depletion analysis methods to analyze systems with multi-batch fuel management schemes. This paper presents three elements of significant novelty. First, the reactivity-equivalent physical transformation (RPT) methodology usually applied in systems with coated particle fuel in cylindrical and spherical geometries was extended to slab geometries. Secondly, based on this newly developed RPT method for slab geometries, a methodology that uses Monte Carlo depletion approaches was further developed to search for the maximum discharge burnup in a multi-batch system by iteratively estimating the beginning of equilibrium cycle composition and sampling different discharge burnups. This iterative equilibrium depletion search (IEDS) method fully defines an equilibrium fuel cycle (keff, power, flux and composition evolutions across space and time), but is computationally demanding, although feasible on single-processor workstations. Finally, an analytical method, the non-linear reactivity model, was developed by expanding the linear reactivity model to include an arbitrary number of higher order terms to extrapolate single-batch depletion results to estimate the maximum discharge burnup and BOEC keff in systems with multi-batch fuel management schemes. Results from this method were benchmarked against equilibrium depletion analysis results using the IEDS.

Cisneros, Anselmo T. [University of California, Berkeley; Ilas, Dan [ORNL

2012-01-01T23:59:59.000Z

118

Neutronics and Depletion Methods for Parametric Studies of Fluoride Salt Cooled High Temperature Reactors with Slab Fuel Geometry and Multi-Batch Fuel Management Schemes  

SciTech Connect

The Advanced High-Temperature Reactor (AHTR) is a 3400 MWth fluoride-salt-cooled high-temperature reactor (FHR) that uses TRISO particle fuel compacted into slabs rather than spherical or cylindrical fuel compacts. Simplified methods are required for parametric design studies such that analyzing the entire feasible design space for an AHTR is tractable. These simplifications include fuel homogenization techniques to increase the speed of neutron transport calculations in depletion analysis and equilibrium depletion analysis methods to analyze systems with multi-batch fuel management schemes. This paper presents three elements of significant novelty. First, the Reactivity-Equivalent Physical Transformation (RPT) methodology usually applied in systems with coated-particle fuel in cylindrical and spherical geometries has been extended to slab geometries. Secondly, based on this newly developed RPT method for slab geometries, a methodology that uses Monte Carlo depletion approaches was further developed to search for the maximum discharge burnup in a multi-batch system by iteratively estimating the beginning of equilibrium cycle (BOEC) composition and sampling different discharge burnups. This Iterative Equilibrium Depletion Search (IEDS) method fully defines an equilibrium fuel cycle (keff, power, flux, and composition evolutions) but is computationally demanding, although feasible on single-processor workstations. Finally, an analytical method, the Non-Linear Reactivity Model, was developed by expanding the linear reactivity model to include an arbitrary number of higher order terms so that single-batch depletion results could be extrapolated to estimate the maximum discharge burnup and BOEC keff in systems with multi-batch fuel management schemes. Results from this method were benchmarked against equilibrium depletion analysis results using the IEDS.

Cisneros, Anselmo T. [University of California, Berkeley; Ilas, Dan [ORNL

2013-01-01T23:59:59.000Z

119

CRAD, Emergency Management - Oak Ridge National Laboratory High...  

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

Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A...

120

CRAD, Emergency Management - Oak Ridge National Laboratory High...  

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

Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor...

Note: This page contains sample records for the topic "reactor aging management" 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

Managing aging effects on dry cask storage systems for extended long-term storage and transportation of used fuel - rev. 0  

Science Conference Proceedings (OSTI)

The cancellation of the Yucca Mountain repository program in the United States raises the prospect of extended long-term storage (i.e., >120 years) and deferred transportation of used fuel at operating and decommissioned nuclear power plant sites. Under U.S. federal regulations contained in Title 10 of the Code of Federal Regulations (CFR) 72.42, the initial license term for an Independent Spent Fuel Storage Installation (ISFSI) must not exceed 40 years from the date of issuance. Licenses may be renewed by the U.S. Nuclear Regulatory Commission (NRC) at the expiration of the license term upon application by the licensee for a period not to exceed 40 years. Application for ISFSI license renewals must include the following: (1) Time-limited aging analyses (TLAAs) that demonstrate that structures, systems, and components (SSCs) important to safety will continue to perform their intended function for the requested period of extended operation; and (2) a description of the aging management program (AMP) for management of issues associated with aging that could adversely affect SSCs important to safety. In addition, the application must also include design bases information as documented in the most recent updated final safety analysis report as required by 10 CFR 72.70. Information contained in previous applications, statements, or reports filed with the Commission under the license may be incorporated by reference provided that those references are clear and specific. The NRC has recently issued the Standard Review Plan (SRP) for renewal of used-fuel dry cask storage system (DCSS) licenses and Certificates of Compliance (CoCs), NUREG-1927, under which NRC may renew a specific license or a CoC for a term not to exceed 40 years. Both the license and the CoC renewal applications must contain revised technical requirements and operating conditions (fuel storage, surveillance and maintenance, and other requirements) for the ISFSI and DCSS that address aging effects that could affect the safe storage of the used fuel. The information contained in the license and CoC renewal applications will require NRC review to verify that the aging effects on the SSCs in DCSSs/ ISFSIs are adequately managed for the period of extended operation. To date, all of the ISFSIs located across the United States with more than 1,500 dry casks loaded with used fuel have initial license terms of 20 years; three ISFSIs (Surry, H.B. Robinson and Oconee) have received their renewed licenses for 20 years, and two other ISFSIs (Calvert Cliffs and Prairie Island) have applied for license renewal for 40 years. This report examines issues related to managing aging effects on the SSCs in DCSSs/ISFSIs for extended long-term storage and transportation of used fuels, following an approach similar to that of the Generic Aging Lessons Learned (GALL) report, NUREG-1801, for the aging management and license renewal of nuclear power plants. The report contains five chapters and an appendix on quality assurance for aging management programs for used-fuel dry storage systems. Chapter I of the report provides an overview of the ISFSI license renewal process based on 10 CFR 72 and the guidance provided in NUREG-1927. Chapter II contains definitions and terms for structures and components in DCSSs, materials, environments, aging effects, and aging mechanisms. Chapter III and Chapter IV contain generic TLAAs and AMPs, respectively, that have been developed for managing aging effects on the SSCs important to safety in the dry cask storage system designs described in Chapter V. The summary descriptions and tabulations of evaluations of AMPs and TLAAs for the SSCs that are important to safety in Chapter V include DCSS designs (i.e., NUHOMS{reg_sign}, HI-STORM 100, Transnuclear (TN) metal cask, NAC International S/T storage cask, ventilated storage cask (VSC-24), and the Westinghouse MC-10 metal dry storage cask) that have been and continue to be used by utilities across the country for dry storage of used fuel to date. The goal of this report is to help establish the technical

Chopra, O.K.; Diercks, D.; Fabian, R.; Ma, D.; Shah, V.; Tam, S.W.; Liu, Y. (Decision and Information Sciences); ( EVS); ( NE)

2012-07-06T23:59:59.000Z

122

Managing aging effects on dry cask storage systems for extended long-term storage and transportation of used fuel - rev. 0  

SciTech Connect

The cancellation of the Yucca Mountain repository program in the United States raises the prospect of extended long-term storage (i.e., >120 years) and deferred transportation of used fuel at operating and decommissioned nuclear power plant sites. Under U.S. federal regulations contained in Title 10 of the Code of Federal Regulations (CFR) 72.42, the initial license term for an Independent Spent Fuel Storage Installation (ISFSI) must not exceed 40 years from the date of issuance. Licenses may be renewed by the U.S. Nuclear Regulatory Commission (NRC) at the expiration of the license term upon application by the licensee for a period not to exceed 40 years. Application for ISFSI license renewals must include the following: (1) Time-limited aging analyses (TLAAs) that demonstrate that structures, systems, and components (SSCs) important to safety will continue to perform their intended function for the requested period of extended operation; and (2) a description of the aging management program (AMP) for management of issues associated with aging that could adversely affect SSCs important to safety. In addition, the application must also include design bases information as documented in the most recent updated final safety analysis report as required by 10 CFR 72.70. Information contained in previous applications, statements, or reports filed with the Commission under the license may be incorporated by reference provided that those references are clear and specific. The NRC has recently issued the Standard Review Plan (SRP) for renewal of used-fuel dry cask storage system (DCSS) licenses and Certificates of Compliance (CoCs), NUREG-1927, under which NRC may renew a specific license or a CoC for a term not to exceed 40 years. Both the license and the CoC renewal applications must contain revised technical requirements and operating conditions (fuel storage, surveillance and maintenance, and other requirements) for the ISFSI and DCSS that address aging effects that could affect the safe storage of the used fuel. The information contained in the license and CoC renewal applications will require NRC review to verify that the aging effects on the SSCs in DCSSs/ ISFSIs are adequately managed for the period of extended operation. To date, all of the ISFSIs located across the United States with more than 1,500 dry casks loaded with used fuel have initial license terms of 20 years; three ISFSIs (Surry, H.B. Robinson and Oconee) have received their renewed licenses for 20 years, and two other ISFSIs (Calvert Cliffs and Prairie Island) have applied for license renewal for 40 years. This report examines issues related to managing aging effects on the SSCs in DCSSs/ISFSIs for extended long-term storage and transportation of used fuels, following an approach similar to that of the Generic Aging Lessons Learned (GALL) report, NUREG-1801, for the aging management and license renewal of nuclear power plants. The report contains five chapters and an appendix on quality assurance for aging management programs for used-fuel dry storage systems. Chapter I of the report provides an overview of the ISFSI license renewal process based on 10 CFR 72 and the guidance provided in NUREG-1927. Chapter II contains definitions and terms for structures and components in DCSSs, materials, environments, aging effects, and aging mechanisms. Chapter III and Chapter IV contain generic TLAAs and AMPs, respectively, that have been developed for managing aging effects on the SSCs important to safety in the dry cask storage system designs described in Chapter V. The summary descriptions and tabulations of evaluations of AMPs and TLAAs for the SSCs that are important to safety in Chapter V include DCSS designs (i.e., NUHOMS{reg_sign}, HI-STORM 100, Transnuclear (TN) metal cask, NAC International S/T storage cask, ventilated storage cask (VSC-24), and the Westinghouse MC-10 metal dry storage cask) that have been and continue to be used by utilities across the country for dry storage of used fuel to date. The goal of this report is to help establish the technical

Chopra, O.K.; Diercks, D.; Fabian, R.; Ma, D.; Shah, V.; Tam, S.W.; Liu, Y. (Decision and Information Sciences); ( EVS); ( NE)

2012-07-06T23:59:59.000Z

123

AECL/US INERI - Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Power Reactors -- Fuel Requirements and Down-Select Report  

Science Conference Proceedings (OSTI)

The U.S. Advanced Fuel Cycle Program and the Atomic Energy Canada Ltd (AECL) seek to develop and demonstrate the technologies needed to minimize the overall Pu and minor actinides present in the light water reactor (LWR) nuclear fuel cycles. It is proposed to reuse the Pu from LWR spent fuel both for the energy it contains and to decrease the hazard and proliferation impact resulting from storage of the Pu and minor actinides. The use of fuel compositions with a combination of U and Pu oxide (MOX) has been proposed as a way to recycle Pu and/or minor actinides in LWRs. It has also been proposed to replace the fertile U{sup 238} matrix of MOX with a fertile-free matrix (IMF) to reduce the production of Pu{sup 239} in the fuel system. It is important to demonstrate the performance of these fuels with the appropriate mixture of isotopes and determine what impact there might be from trace elements or contaminants. Previous work has already been done to look at weapons-grade (WG) Pu in the MOX configuration [1][2] and the reactor-grade (RG) Pu in a MOX configuration including small (4000 ppm additions of Neptunium). This program will add to the existing database by developing a wide variety of MOX fuel compositions along with new fuel compositions called inert-matrix fuel (IMF). The goal of this program is to determine the general fabrication and irradiation behavior of the proposed IMF fuel compositions. Successful performance of these compositions will lead to further selection and development of IMF for use in LWRs. This experiment will also test various inert matrix material compositions with and without quantities of the minor actinides Americium and Neptunium to determine feasibility of incorporation into the fuel matrices for destruction. There is interest in the U.S. and world-wide in the investigation of IMF (inert matrix fuels) for scenarios involving stabilization or burn down of plutonium in the fleet of existing commercial power reactors. IMF offer the potential advantage for more efficient destruction of plutonium and minor actinides (MA) relative to MOX fuel. Greater efficiency in plutonium reduction results in greater flexibility in managing plutonium inventories and in developing strategies for disposition of MA, as well as a potential for fuel cycle cost savings. Because fabrication of plutonium-bearing (and MA-bearing) fuel is expensive relative to UO{sub 2} in terms of both capital and production, cost benefit can be realized through a reduction in the number of plutonium-bearing elements required for a given burn rate. In addition, the choice of matrix material may be manipulated either to facilitate fuel recycling or to make plutonium recovery extremely difficult. In addition to plutonium/actinide management, an inert matrix fuel having high thermal conductivity may have operational and safety benefits; lower fuel temperatures could be used to increase operating and safety margins, uprate reactor power, or a combination of both. The CANDU reactor offers flexibility in plutonium management and MA burning by virtue of online refueling, a simple bundle design, and good neutron economy. A full core of inert matrix fuel containing either plutonium or a plutonium-actinide mix can be utilized, with plutonium destruction efficiencies greater than 90%, and high (>60%) actinide destruction efficiencies. The Advanced CANDU Reactor (ACR) could allow additional possibilities in the design of an IMF bundle, since the tighter lattice pitch and light-water coolant reduce or eliminate the need to suppress coolant void reactivity, allowing the center region of the bundle to include additional fissile material and to improve actinide burning. The ACR would provide flexibility for management of plutonium and MA from the existing LWR fleet, and would be complementary to the AFCI program in the U.S. Many of the fundamental principles concerning the use of IMF are nearly identical in LWRs and the ACR, including fuel/coolant compatibility, fuel fabrication, and fuel irradiation behavior. In addition, the U.S. and Canada both

William Carmack; Randy D. Lee; Pavel Medvedev; Mitch Meyer; Michael Todosow; Holly B. Hamilton; Juan Nino; Simon Philpot; James Tulenko

2005-06-01T23:59:59.000Z

124

Plutonium Recycling in Light Water Reactors at Framatome ANP: Status and Trends  

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Use of Alternate Fuels in Light Water Reactors

Dieter Porsch; Walter Stach; Pascal Charmensat; Michel Pasquet

125

Nuclear Reactors  

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

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

126

BNL Environmental Management Directorate  

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

Surface Projects Brookhaven Graphite Research Reactor High Flux Beam Reactor Waste Management Division OU IV Five-Year Review The five-year review of the Operable Unit IV...

127

Long-Term Operations Program: Assessment of Research and Development Supporting Aging Management Programs for Long-Term Operation  

Science Conference Proceedings (OSTI)

Presently, U.S. nuclear utilities and the Nuclear Regulatory Commission (NRC) are discussing a second round of plant license renewals, termed subsequent license renewal (SLR). For the U.S. plants this could mean extending a plant’s operating license from 60 years to 80. To achieve safe and reliable operation for such an extended period will require a comprehensive technical understanding of aging degradation effects on nuclear plant systems, structures, and components (SSCs). EPRI ...

2013-08-29T23:59:59.000Z

128

Twenty-First Water Reactor Safety Information Meeting. Volume 3, Primary system integrity; Aging research, products and applications; Structural and seismic engineering; Seismology and geology: Proceedings  

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. Selected papers were indexed separately for inclusion in the Energy Science and Technology Database.

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

1994-04-01T23:59:59.000Z

129

Light Water Reactor Sustainability Program Technical Basis Guide Describing How to Perform Safety Margin Configuration Risk Management  

SciTech Connect

The INL has carried out a demonstration of the RISMC approach for the purpose of configuration risk management. We have shown how improved accuracy and realism can be achieved by simulating changes in risk – as a function of different configurations – in order to determine safety margins as the plant is modified. We described the various technical issues that play a role in these configuration-based calculations with the intent that future applications can take advantage of the analysis benefits while avoiding some of the technical pitfalls that are found for these types of calculations. Specific recommendations have been provided on a variety of topics aimed at improving the safety margin analysis and strengthening the technical basis behind the analysis process.

Curtis Smith; James Knudsen; Bentley Harwood

2013-08-01T23:59:59.000Z

130

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

131

A 48-month extended fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor  

Science Conference Proceedings (OSTI)

The B and W mPower{sup TM} reactor is a small, rail-shippable pressurized water reactor (PWR) with an integral once-through steam generator and an electric power output of 150 MW, which is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height, but otherwise standard, PWR assemblies with the familiar 17 x 17 fuel rod array on a 21.5 cm inter-assembly pitch. The B and W mPower core design and cycle management plan, which were performed using the Studsvik core design code suite, follow the pattern of a typical nuclear reactor fuel cycle design and analysis performed by most nuclear fuel management organizations, such as fuel vendors and utilities. However, B and W is offering a core loading and cycle management plan for four years of continuous power operations without refueling and without the hurdles of chemical shim. (authors)

Erighin, M. A. [Babcock and Wilcox Company, 109 Ramsey Place, Lynchburg, VA 24502 (United States)

2012-07-01T23:59:59.000Z

132

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Treshow, M.

1961-09-01T23:59:59.000Z

133

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

Daniels, F.

1959-10-27T23:59:59.000Z

134

CONVECTION REACTOR  

DOE Patents (OSTI)

An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

Hammond, R.P.; King, L.D.P.

1960-03-22T23:59:59.000Z

135

CRC handbook of nuclear reactors calculations. Vol. II  

Science Conference Proceedings (OSTI)

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

Ronen, Y.

1986-01-01T23:59:59.000Z

136

Management  

E-Print Network (OSTI)

Research has indicated that, depending upon driver and passenger characteristics, passengers can have either a positive or negative influence upon driver behaviour. In conclusion to a recent study investigating the roles that passengers can play to influence, positively and negatively, driver behaviour, Regan and Mitsopoulos (2001) recommended, among other things, that the principles of Crew Resource Management (CRM) training may increase passengers ’ ability to positively influence driver behaviour and also drivers ’ ability to accept constructive feedback. The present study investigated the potential application of CRM training within young driver training in the Australian Capital Territory (ACT). This involved a literature review, an analysis of the differences between the driving and aviation domains, an analysis of the team-based activities and the knowledge, skills and attitudes required during driving to perform those activities, consultation with CRM experts from the aviation and medicine domains and the conduct of six focus groups involving young learner drivers, provisional licence drivers and course teachers. The findings indicate that CRM training as part of young driver training in the ACT is a viable concept to pursue. The application of CRM training within young driver training has potential to significantly enhance the positive and reduce the negative effects of passengers on young driver behaviour, and thus the safety of young drivers and passengers alike. The outcomes of this study formed the basis for a set of recommendations for the development of a young driver CRM training program in the ACT.

Young Drivers; Eve Mitsopoulos; Michael Regan; Janet Anderson; Paul Salmon; Jessica Edquist; Ii Report Documentation Page

2005-01-01T23:59:59.000Z

137

CRAD, Configuration Management - Oak Ridge National Laboratory...  

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

Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Configuration Management - Oak Ridge National Laboratory High Flux Isotope...

138

ME 361E Nuclear Reactor Engineering ABET EC2000 syllabus  

E-Print Network (OSTI)

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

Ben-Yakar, Adela

139

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

140

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

Fraas, A.P.; Mills, C.B.

1961-11-21T23:59:59.000Z

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


141

REACTOR COOLING  

DOE Patents (OSTI)

A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

Quackenbush, C.F.

1959-09-29T23:59:59.000Z

142

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

Wigner, E.P.

1958-04-22T23:59:59.000Z

143

Nuclear power plant Generic Aging Lessons Learned (GALL). Appendix B  

SciTech Connect

The purpose of this generic aging lessons learned (GALL) review is to provide a systematic review of plant aging information in order to assess materials and component aging issues related to continued operation and license renewal of operating reactors. Literature on mechanical, structural, and thermal-hydraulic components and systems reviewed consisted of 97 Nuclear Plant Aging Research (NPAR) reports, 23 NRC Generic Letters, 154 Information Notices, 29 Licensee Event Reports (LERs), 4 Bulletins, and 9 Nuclear Management and Resources Council Industry Reports (NUMARC IRs) and literature on electrical components and systems reviewed consisted of 66 NPAR reports, 8 NRC Generic Letters, 111 Information Notices, 53 LERs, 1 Bulletin, and 1 NUMARC IR. More than 550 documents were reviewed. The results of these reviews were systematized using a standardized GALL tabular format and standardized definitions of aging-related degradation mechanisms and effects. The tables are included in volume s 1 and 2 of this report. A computerized data base has also been developed for all review tables and can be used to expedite the search for desired information on structures, components, and relevant aging effects. A survey of the GALL tables reveals that all ongoing significant component aging issues are currently being addressed by the regulatory process. However, the aging of what are termed passive components has been highlighted for continued scrutiny. This report consists of Volume 2, which consists of the GALL literature review tables for the NUMARC Industry Reports reviewed for the report.

Kasza, K.E.; Diercks, D.R.; Holland, J.W.; Choi, S.U. [and others

1996-12-01T23:59:59.000Z

144

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

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

1958-12-01T23:59:59.000Z

145

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

National Nuclear Security Administration (NNSA)

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

146

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

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

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

147

Reactor Materials  

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

The reactor materials crosscut effort will enable the development of innovative and revolutionary materials and provide broad-based, modern materials science that will benefit all four DOE-NE...

148

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

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

1962-10-23T23:59:59.000Z

149

NEUTRONIC REACTORS  

DOE Patents (OSTI)

A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

Wigner, E.P.

1960-11-22T23:59:59.000Z

150

REACTOR SHIELD  

DOE Patents (OSTI)

Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

1959-02-17T23:59:59.000Z

151

Advanced Light Water Reactor - Boiling Water Reactor Degradation Matrix (ALWR BWR DM), Revision 0  

Science Conference Proceedings (OSTI)

The advanced light water reactor–boiling water reactor degradation matrix (ALWR BWR DM) is an essential piece of the Electric Power Research Institute’s (EPRI’s) Advanced Nuclear Technology (ANT) materials management matrix initiative for advanced LWR designs. The materials management matrix provides a tool to assist the industry in proactive identification and consideration of materials issues as well as mitigation and management opportunities from the design phase, through component fabrication and pla...

2009-08-25T23:59:59.000Z

152

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Grebe, J.J.

1959-07-14T23:59:59.000Z

153

Light Water Reactor Sustainability Program - Non-Destructive...  

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

Remaining Useful Life of Aging Cables in Nuclear Power Plants Light Water Reactor Sustainability Program - Non-Destructive Evaluation R&D Roadmap for Determining Remaining Useful...

154

EM's Top Official Celebrates 'Cocooning' of Reactor Dedicated by  

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

Top Official Celebrates 'Cocooning' of Reactor Dedicated Top Official Celebrates 'Cocooning' of Reactor Dedicated by President Kennedy Decades Ago EM's Top Official Celebrates 'Cocooning' of Reactor Dedicated by President Kennedy Decades Ago June 1, 2012 - 12:00pm Addthis Senior Advisor for Environmental Management David Huizenga speaks during an event announcing the completion of work to place N Reactor in safe storage. Senior Advisor for Environmental Management David Huizenga speaks during an event announcing the completion of work to place N Reactor in safe storage. An aerial photo shows the N Reactor complex just before work to place the reactor in safe storage, or "cocooning," was completed. An aerial photo shows the N Reactor complex just before work to place the reactor in safe storage, or "cocooning," was completed.

155

Overview of the US Department of Energy Light Water Reactor Sustainability Program  

Science Conference Proceedings (OSTI)

The US Department of Energy Light Water Reactor Sustainability Program is focused on the long-term operation of US commercial power plants. It encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper gives an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables.

K. A. McCarthy; D. L. Williams; R. Reister

2012-05-01T23:59:59.000Z

156

Steam Generator Management Program: Pressurized Water Reactor Generic Tube Degradation Predictions: Recirculating Steam Generators with Alloy 600TT, Alloy 690TT, and Alloy 800NG Tubing  

Science Conference Proceedings (OSTI)

Mill-annealed Alloy 600 heat transfer tubing in pressurized water reactor (PWR) steam generators (SGs) has experienced numerous modes of degradation. This report describes predictive models for determining expected tube degradation in recirculating steam generators with Alloy 600TT, Alloy 690TT, and Alloy 800NG tubing. Predictions are based on operating experience with similar designs and use improvement factors to characterize benefits resulting from SG design and material ...

2013-12-17T23:59:59.000Z

157

Advances in reactor physics education: Visualization of reactor parameters  

Science Conference Proceedings (OSTI)

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

Snoj, L.; Kromar, M.; Zerovnik, G. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

2012-07-01T23:59:59.000Z

158

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

159

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...

160

NEUTRONIC REACTOR  

DOE Patents (OSTI)

This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

1958-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor aging management" 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

2012 Awards for Project Management | Department of Energy  

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

Daya Bay Reactor Neutrino Detector Project The Office of Environmental Management's Idaho Nuclear facility Decontamination & Decommissioning Project More Documents & Publications...

162

Evaluation of Thermal Fatigue Effects on Systems Requiring Aging Management Review for License Renewal for the Calvert Cliffs Nuclear Power Plant  

Science Conference Proceedings (OSTI)

In recent years, the nuclear power industry has devoted significant attention to metal fatigue and its impact on the design qualification and serviceability of operating components. This study provides a pilot plant demonstration of the current industry technical position on fatigue evaluation for license renewal, specifically addressing reactor water environmental effects. It also develops a technical evaluation method for determining fatigue life adequacy in a feedwater piping system, a pressurizer sur...

1998-02-25T23:59:59.000Z

163

POWER REACTOR  

DOE Patents (OSTI)

A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

Zinn, W.H.

1958-07-01T23:59:59.000Z

164

REACTOR CONTROL  

DOE Patents (OSTI)

A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

Fortescue, P.; Nicoll, D.

1962-04-24T23:59:59.000Z

165

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Christy, R.F.

1958-07-15T23:59:59.000Z

166

Catalytic reactor  

DOE Patents (OSTI)

A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

2009-03-10T23:59:59.000Z

167

NEUTRONIC REACTORS  

DOE Patents (OSTI)

A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

Wigner, E.P.; Young, G.J.

1958-10-14T23:59:59.000Z

168

Overview of Reactor and Nuclear  

E-Print Network (OSTI)

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

169

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Young, G.

1963-01-01T23:59:59.000Z

170

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

Wigner, E.P.; Weinberg, A.W.; Young, G.J.

1958-04-15T23:59:59.000Z

171

Power Burst Facility (PBF) Reactor Reactor Decommissioning  

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

Reactor Decommissioning Click here to view Click here to view Reactor Decommissioning Click on an image to enlarge A crane removes the reactor vessel from the Power Burst Facility...

172

Plant Support Engineering: Aging Management Program Development Guidance for AC and DC Low-Voltage Power Cable Systems for Nuclear P ower Plants  

Science Conference Proceedings (OSTI)

Regulatory and management concern regarding the reliability of low-voltage power cable systems at nuclear plants has been increasing for the past 5–10 years. The staff of the United States Nuclear Regulatory Commission are concerned that wetted (up to and including submergence) low-voltage power cable circuits may be degrading to the point at which multiple cable circuits may fail when called on to perform functions affecting safety. Utility managers are concerned that cables may fail, causing adverse sa...

2010-06-02T23:59:59.000Z

173

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A reactor is described comprising a plurality of horizontal trays containing a solution of a fissionable material, the trays being sleeved on a vertical tube which contains a vertically-reciprocable control rod, a gas-tight chamber enclosing the trays, and means for conducting vaporized moderator from the chamber and for replacing vaporized moderator in the trays. (AEC)

Wigner, E.P.

1962-12-25T23:59:59.000Z

174

Neutronic reactor  

DOE Patents (OSTI)

A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

Wende, Charles W. J. (West Chester, PA)

1976-08-17T23:59:59.000Z

175

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Post, R.G.

1963-05-01T23:59:59.000Z

176

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Starr, C.

1963-01-01T23:59:59.000Z

177

NEUTRONIC REACTOR  

DOE Patents (OSTI)

BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

1959-10-27T23:59:59.000Z

178

NEUTRONIC REACTORS  

DOE Patents (OSTI)

The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

Anderson, H.L.

1958-10-01T23:59:59.000Z

179

Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system  

Science Conference Proceedings (OSTI)

Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

Harto, Andang Widi [Engineering Physics Department, Faculty of Engineering, Gadjah Mada University (Indonesia)

2012-06-06T23:59:59.000Z

180

Aging gene  

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

Aging gene Name: Linda S Martinez Location: NA Country: NA Date: NA Question: Have the aging gene or genes been located on the human chromosomes, and, if yes, will removing that...

Note: This page contains sample records for the topic "reactor aging management" 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

Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel  

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Increased Enrichment/High Burnup and Light Water Reactor Fuel Cycle Optimization

B. T. Rearden; W. J. Anderson; G. A. Harms

182

Life Cycle Management Planning at Wolf Creek Generating Station: EDG, Main Steam, and Feedwater Isolation Valves, and Reactor Protec tion System  

Science Conference Proceedings (OSTI)

As the electric power industry becomes more competitive, life cycle management (LCM) of systems, structures, and components (SSCs) becomes very important to keep plants economically viable throughout their remaining licensed operating terms (either 40-year or 60-year terms). This report provides the industry with lessons learned from applying EPRI's LCM planning process to three SSCs at Wolf Creek Generating Station.

2001-12-19T23:59:59.000Z

183

Materials Reliability Program: Reactor Pressure Vessel Integrity Primer (MRP-278)  

Science Conference Proceedings (OSTI)

This primer is based on two earlier Electric Power Research Institute (EPRI) reports: Reactor Vessel Embrittlement Management Handbook: A Handbook for Managing Reactor Vessel Embrittlement and Vessel Integrity (TR-101975-T2) and Primer: Fracture Mechanics in the Nuclear Power Industry (NP-5792-SR, Rev. 1). The information in those earlier reports has been updated extensively and focuses on todays reactor pressure vessel (RPV) embrittlement, integrity, and plant license renewal issues. This RPV integrity ...

2010-06-09T23:59:59.000Z

184

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

185

REACTOR UNLOADING  

DOE Patents (OSTI)

This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

Leverett, M.C.

1958-02-18T23:59:59.000Z

186

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Treshow, M.

1958-08-19T23:59:59.000Z

187

Advanced Nuclear Technology: EPRI Materials Management Matrix Project—Toshiba Advanced Boiling Water Reactor Materials Managem ent Table Report, Revision 0  

Science Conference Proceedings (OSTI)

Experience gained through years of operating nuclear plants has shown that materials performance issues can be a significant concern related to economic and safe long-term plant operations. Although concerns remain, industry efforts to address materials performance issues at operating plants have led to several important advances in both the underlying scientific understanding of materials degradation and the implementation of practical mitigation and management technologies. The Electric Power Research...

2010-02-09T23:59:59.000Z

188

Neutronic reactor  

DOE Patents (OSTI)

A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

Lewis, Warren R. (Richland, WA)

1978-05-30T23:59:59.000Z

189

NUCLEAR REACTORS  

DOE Patents (OSTI)

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

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

1961-12-01T23:59:59.000Z

190

REACTOR CONTROL  

DOE Patents (OSTI)

This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

Ruano, W.J.

1957-12-10T23:59:59.000Z

191

F Reactor Area Cleanup Complete | Department of Energy  

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

F Reactor Area Cleanup Complete F Reactor Area Cleanup Complete F Reactor Area Cleanup Complete September 19, 2012 - 12:00pm Addthis Media Contact Cameron Hardy, DOE Cameron.Hardy@rl.doe.gov 509-376-5365 RICHLAND, Wash. - U.S. Department of Energy (DOE) contractors have cleaned up the F Reactor Area, the first reactor area at the Hanford Site in southeastern Washington state to be fully remediated. While six of Hanford's nine plutonium production reactors have been sealed up, or cocooned, the F Reactor Area is the first to have all of its associated buildings and waste sites cleaned up in addition to having its reactor sealed up. "The cleanup of the F Reactor Area shows the tremendous progress workers are making along Hanford's River Corridor," said Dave Huizenga, Senior Advisor for the DOE Office of Environmental Management. "The River

192

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

193

CHF Enhancement by Vessel Coating for External Reactor Vessel Cooling  

SciTech Connect

In-vessel retention (IVR) is a key severe accident management (SAM) strategy that has been adopted by some operating nuclear power plants and advanced light water reactors (ALWRs). One viable means for IVR is the method of external reactor vessel cooling (ERVC) by flooding of the reactor cavity during a severe accident. As part of a joint Korean – United States International Nuclear Energy Research Initiative (K-INERI), an experimental study has been conducted to investigate the viability of using an appropriate vessel coating to enhance the critical heat flux (CHF) limits during ERVC. Toward this end, transient quenching and steady-state boiling experiments were performed in the SBLB (Subscale Boundary Layer Boiling) facility at Penn State using test vessels with micro-porous aluminum coatings. Local boiling curves and CHF limits were obtained in these experiments. When compared to the corresponding data without coatings, substantial enhancement in the local CHF limits for the case with surface coatings was observed. Results of the steady state boiling experiments showed that micro-porous aluminum coatings were very durable. Even after many cycles of steady state boiling, the vessel coatings remained rather intact, with no apparent changes in color or structure. Moreover, the heat transfer performance of the coatings was found to be highly desirable with an appreciable CHF enhancement in all locations on the vessel outer surface but with very little effect of aging.

Fan-Bill Cheung; Joy L. Rempe

2004-06-01T23:59:59.000Z

194

Aging gauge  

DOE Patents (OSTI)

An aging gauge comprising a container having a fixed or a variable sized t opening with a cap which can be opened to control the sublimation rate of a thermally sublimational material contained within the container. In use, the aging gauge is stored with an item to determine total heat the item is subjected to and also the maximum temperature to which the item has been exposed. The aging gauge container contains a thermally sublimational material such as naphthalene or similar material which has a low sublimation rate over the temperature range from about 70.degree. F. to about 160.degree. F. The aging products determined by analyses of a like item aged along with the aging gauge for which the sublimation amount is determined is employed to establish a calibration curve for future aging evaluation. The aging gauge is provided with a means for determining the maximum temperature exposure (i.e., a thermally indicating material which gives an irreversible color change, Thermocolor pigment). Because of the relationship of doubling reaction rates for increases of 10.degree. C., equivalency of item used in accelerated aging evaluation can be obtained by referring to a calibration curve depicting storage temperature on the abscissa scale and multiplier on the ordinate scale.

Betts, Robert E. (Huntsville, AL); Crawford, John F. (Huntsville, AL)

1989-01-01T23:59:59.000Z

195

Nuclear reactor  

DOE Patents (OSTI)

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

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

1977-01-01T23:59:59.000Z

196

ELECTRONUCLEAR REACTOR  

DOE Patents (OSTI)

An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.

Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.

1960-04-19T23:59:59.000Z

197

Photocatalytic reactor  

DOE Patents (OSTI)

A photocatalytic reactor for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane.

Bischoff, Brian L. (Knoxville, TN); Fain, Douglas E. (Oak Ridge, TN); Stockdale, John A. D. (Knoxville, TN)

1999-01-01T23:59:59.000Z

198

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

199

CONTROL MEANS FOR REACTOR  

DOE Patents (OSTI)

An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

Manley, J.H.

1961-06-27T23:59:59.000Z

200

Materials Reliability Program: Reactor Pressure Vessel Integrity Training Module (MRP-286)  

Science Conference Proceedings (OSTI)

For many reactor pressure vessels, embrittlement is the primary concern in ensuring continued safe operation. The shutdown of the Yankee Rowe plant, which occurred because of uncertainties related to embrittlement of the vessel, demonstrated the importance of adequately addressing embrittlement issues. Managing embrittlement involves the integration, management, and implementation of diverse technical, regulatory, planning, and economic activities. Reactor vessel embrittlement management is an essential ...

2010-11-22T23:59:59.000Z

Note: This page contains sample records for the topic "reactor aging management" 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

Nuclear Reactor Accidents  

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

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

202

Principles of Reactor Physics  

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

Nuclear Reactor Physics M A Smith Argonne National Laboratory Nuclear Engineering Division Phone: 630-252-9747, Email: masmith@anl.gov Abstract: Nuclear reactor physics deals with...

203

Ground test facility for nuclear testing of space reactor subsystems  

SciTech Connect

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

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

1985-01-01T23:59:59.000Z

204

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

Daniels, F.

1962-12-18T23:59:59.000Z

205

Light Water Reactor Sustainability (LWRS) Program | Department of Energy  

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

Light Water Reactor Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program The Light Water Reactor Sustainability (LWRS) Program is developing the scientific basis to extend existing nuclear power plant operating life beyond the current 60-year licensing period and ensure long-term reliability, productivity, safety, and security. The program is conducted in collaboration with national laboratories, universities, industry, and international partners. Idaho National Laboratory serves as the Technical Integration Office and coordinates the research and development (R&D) projects in the following pathways: Materials Aging and Degradation Assessment, Advanced Instrumentation, Information, and Control Systems

206

Reactor and method of operation  

DOE Patents (OSTI)

A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

Wheeler, John A. (Princeton, NJ)

1976-08-10T23:59:59.000Z

207

Emergency Management Criteria & Review Approach Documents | Department of  

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

Emergency Management Criteria & Review Approach Documents Emergency Management Criteria & Review Approach Documents Emergency Management Criteria & Review Approach Documents Documents Available for Download CRAD, Emergency Management - Idaho Accelerated Retrieval Project Phase II CRAD, Emergency Management - Los Alamos National Laboratory TA 55 SST Facility CRAD, Emergency Management - Los Alamos National Laboratory Waste Characterization, Reduction, and Repackaging Facility CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Emergency Management - Office of River Protection K Basin Sludge Waste System CRAD, Emergency Management - Y-12 Enriched Uranium Operations Oxide

208

Reactor safety method  

DOE Patents (OSTI)

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

Vachon, Lawrence J. (Clairton, PA)

1980-03-11T23:59:59.000Z

209

NEUTRONIC REACTOR MANIPULATING DEVICE  

DOE Patents (OSTI)

A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

Ohlinger, L.A.

1962-08-01T23:59:59.000Z

210

Sandia LDRD-funded project looks at advanced materials for reactors...  

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

LDRD-funded project looks at advanced materials for reactors | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the...

211

An Engineering Test Reactor  

SciTech Connect

A relatively inexpensive reactor for the specific purpose of testing a sub-critical portion of another reactor under conditions that would exist during actual operation is discussed. It is concluded that an engineering tool for reactor development work that bridges the present gap between exponential and criticality experiments and the actual full scale operating reactor is feasible. An example of such a test reactor which would not entail development effort to ut into operation is depicted.

Fahrner, T.; Stoker, R.L.; Thomson, A.S.

1951-03-16T23:59:59.000Z

212

Materials and Fuels for the Current and Advanced Nuclear Reactors II  

Science Conference Proceedings (OSTI)

Lifetime extension of reactors - nuclear materials aging, degradation and others ... A Rate-Theory Approach to Irradiation Damage Modeling with Random ...

213

Materials and Fuels for the Current and Advanced Nuclear Reactors III  

Science Conference Proceedings (OSTI)

Jul 15, 2013 ... Lifetime extension of reactors - nuclear materials aging, degradation and ... A Theoretical Model of Corrosion Rate Distribution in Liquid LBE ...

214

SRS Seals Access to P and R Reactors, Marking End to Nearly 60-Years of  

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

Seals Access to P and R Reactors, Marking End to Nearly Seals Access to P and R Reactors, Marking End to Nearly 60-Years of History: American Recovery and Reinvestment Act Project Progress SRS Seals Access to P and R Reactors, Marking End to Nearly 60-Years of History: American Recovery and Reinvestment Act Project Progress June 28, 2011 - 12:00pm Addthis Marc Sharpe, who was a senior reactor operator at P Reactor in the mid-1980s, (left) and Dr. David Moody, U.S. Department of Energy-Savannah River Operation Office Manager, were the last people to leave P Reactor, just before the opening is welded shut. Marc Sharpe, who was a senior reactor operator at P Reactor in the mid-1980s, (left) and Dr. David Moody, U.S. Department of Energy-Savannah River Operation Office Manager, were the last people to leave P Reactor,

215

Preliminary Transportation, Aging and Disposal Canister System Performance Specification  

SciTech Connect

This document provides specifications for selected system components of the Transportation, Aging and Disposal (TAD) canister-based system. A list of system specified components and ancillary components are included in Section 1.2. The TAD canister, in conjunction with specialized overpacks will accomplish a number of functions in the management and disposal of spent nuclear fuel. Some of these functions will be accomplished at purchaser sites where commercial spent nuclear fuel (CSNF) is stored, and some will be performed within the Office of Civilian Radioactive Waste Management (OCRWM) transportation and disposal system. This document contains only those requirements unique to applications within Department of Energy's (DOE's) system. DOE recognizes that TAD canisters may have to perform similar functions at purchaser sites. Requirements to meet reactor functions, such as on-site dry storage, handling, and loading for transportation, are expected to be similar to commercially available canister-based systems. This document is intended to be referenced in the license application for the Monitored Geologic Repository (MGR). As such, the requirements cited herein are needed for TAD system use in OCRWM's disposal system. This document contains specifications for the TAD canister, transportation overpack and aging overpack. The remaining components and equipment that are unique to the OCRWM system or for similar purchaser applications will be supplied by others.

C.A Kouts

2006-11-22T23:59:59.000Z

216

Reactor Pressure Vessel Task of Light Water Reactor Sustainability...  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure...

217

CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor |  

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

Reactor Reactor CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Training Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. RADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor

218

CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor  

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

Reactor Reactor CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Maintenance Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Safety Basis - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor

219

Membrane reactor advantages for methanol reforming and similar reactions  

Science Conference Proceedings (OSTI)

Membrane reactors achieve efficiencies by combining in one unit a reactor that generates a product with a semipermeable membrane that extracts it. One well-known benefit of this is greater conversion, as removal of a product drives reactions toward completion, but there are several potentially larger advantages that have been largely ignored. Because a membrane reactor tends to limit the partial pressure of the extracted product, it fundamentally changes the way that total pressure in the reactor affects equilibrium conversion. Thus, many gas-phase reactions that are preferentially performed at low pressures in a conventional reactor are found to have maximum conversion at high pressures in a membrane reactor. These higher pressures and reaction conversions allow greatly enhanced product extraction as well. Further, membrane reactors provide unique opportunities for temperature management which have not been discussed previously. These benefits are illustrated for methanol reforming to hydrogen for use with PEM (polymer electrolyte membrane) fuel cells.

Buxbaum, R.E. [REB Research and Consulting Co., Ferndale, MI (United States)

1999-07-01T23:59:59.000Z

220

Development of a reactor engineering workstation at Seabrook station  

SciTech Connect

The reactor engineers at Seabrook station are responsible for supporting plant operation with respect to the current reactor core design. Advanced assembly designs, complex reactor core loading patterns, and emphasis on efficient and safe operation puts a greater demand on the reactor engineer. The traditional use of static data constants and coarse core modeling, in light of the more complex fuel and core designs of today, results in less than optimum monitoring and predicting tools for the reactor engineer. The incorporation of an advanced three-dimensional nodal code with thermal feedbacks and detailed spatial modeling along with the ability to follow current operational history on a state-of-the-art workstation provides the reactor engineer with a dynamic core monitoring and predictive tool. This approach allows for more accurate and efficient completion of the reactor engineer's tasks. Yankee Atomic Electric Company (YAEC) is currently in the process of providing advanced reactor physics nodal methods to the reactor engineers at Seabrook station. The scope of this project is to supply a reactor engineering workstation with a simplified user interface to an advanced nodal core model as part of an on-line core monitor/predictor for standard reactor engineering calculations. It uses the Studsvik Core Management System (CMS), which primarily consists of the CASMO-3 cross-section generating code and the SIMULATE-3 three-dimensional two-group nodal reactor analysis code.

Tremblay, M.A.; Gorski, J.P. (Yankee Atomic Electric Co., Bolton, MA (United States)); Gurney, P.V. (New Hampshire Yankee, Seabrook, NY (United States))

1992-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor aging management" 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

Memorandum on Chemical Reactors and Reactor Hazards  

SciTech Connect

Two important problems in the investigation of reactor hazards are the chemical reactivity of various materials employed in reactor construction and the chracteristics of heat transfer under transient conditions, specifically heat transfer when driven by an exponentially increasing heat source (exp t/T). Although these problems are independent of each other, when studied in relation to reactor hazards they may occur in a closely coupled sequence. For example the onset of a dangerous chemical reactor may be due to structural failure of various reactor components under an exponentially rising heat source originating with a runaway nuclear reactor. For this reason, these two problems should eventually be studied together after an exploratory experimental survey has been made in which they are considered separately.

Mills, M.M.; Pearlman, H.; Ruebsamen, W.; Steele, G., Chrisney, J.

1951-07-05T23:59:59.000Z

222

Civilian Radioactive Waste Management System Requirements Document...  

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

Management System Requirements Document More Documents & Publications Transportation, Aging and Disposal Canister System Performance Specification: Revision 1 FY 2007 Total...

223

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...

224

Cultural (Historical) Resource Management, Environmental Protection  

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

History Homepage History Homepage Accelerators & Detectors Cosmotron AGS Strong-focusing 80" Bubble Chamber The omega-minus 7' Bubble Chamber The charmed baryon NSLS RHIC Reactors Graphite Research Reactor High Flux Beam Reactor Medical Research Reactor Life Sciences Medical breakthroughs Biology research Plant Genetics Other BNL Nobel Prizes The First Video Game? BNL Physics Timeline Camp Upton Historic Images BNL Cultural (Historical) Resource Management Cultural (Historical) Resource Management at Brookhaven National Laboratory Photograph of the remains of WWI training trenches The Environmental Protection Division is responsible for ensuring compliance with historic preservation requirements. The BNL Cultural Resource Management Plan identifies and describes the management plans for of all of BNL's cultural resources. These resources include World War I trenches, Civilian Conservation Corps features, World War II buildings, and historic structures, programs and discoveries associated with high energy physics, research reactors, and other science conducted at the Laboratory.

225

Attrition reactor system  

DOE Patents (OSTI)

A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

1993-01-01T23:59:59.000Z

226

Guidebook to nuclear reactors  

SciTech Connect

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

Nero, A.V. Jr.

1976-05-01T23:59:59.000Z

227

Reactor Sharing Program  

Science Conference Proceedings (OSTI)

Progress achieved at the University of Florida Training Reactor (UFTR) facility through the US Department of Energy's University Reactor Sharing Program is reported for the period of 1991--1992.

Vernetson, W.G.

1993-01-01T23:59:59.000Z

228

Attrition reactor system  

DOE Patents (OSTI)

A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

Scott, C.D.; Davison, B.H.

1993-09-28T23:59:59.000Z

229

NEUTRONIC REACTOR POWER PLANT  

DOE Patents (OSTI)

This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

Metcalf, H.E.

1962-12-25T23:59:59.000Z

230

Compare Activities by Building Age  

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

Activities by Building Age Activities by Building Age Compare Activities by ... Building Age Median Age of Building by Building Type Vacant buildings, retail stores (other than malls), and religious worship buildings tended to be the oldest buildings. Food sales buildings (which were predominantly convenience stores) and outpatient health care buildings were mainly newer buildings. Figure showing median age of building by building type. If you need assistance viewing this page, please call 202-586-8800. Specific questions may be directed to: Joelle Michaels joelle.michaels@eia.doe.gov CBECS Manager Release date: July 24, 2002 Page last modified: May 4, 2009 2:52 PM http://www.eia.gov/consumption/commercial/data/archive/cbecs/pba99/compareage.html If you are having any technical problems with this site, please contact the EIA

231

High solids fermentation reactor  

DOE Patents (OSTI)

A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

Wyman, Charles E. (Lakewood, CO); Grohmann, Karel (Littleton, CO); Himmel, Michael E. (Littleton, CO); Richard, Christopher J. (Lakewood, CO)

1993-01-01T23:59:59.000Z

232

Improved vortex reactor system  

DOE Patents (OSTI)

An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

1995-01-01T23:59:59.000Z

233

FAST NEUTRON REACTOR  

DOE Patents (OSTI)

A reactor comprising fissionable material in concentration sufficiently high so that the average neutron enengy within the reactor is at least 25,000 ev is described. A natural uranium blanket surrounds the reactor, and a moderating reflector surrounds the blanket. The blanket is thick enough to substantially eliminate flow of neutrons from the reflector.

Soodak, H.; Wigner, E.P.

1961-07-25T23:59:59.000Z

234

NUCLEAR REACTOR CONTROL SYSTEM  

DOE Patents (OSTI)

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

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

1959-11-01T23:59:59.000Z

235

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

236

Electric Power Produced from Nuclear Reactor | National Nuclear Security  

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

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

237

Managing Aging Effects on Dry Cask Storage  

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

... III.4-5 III.5 Time-Dependent Degradation of Radiation-Shielding Materials ... III.5-1 III.5.1 Description of the...

238

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

239

Nuclear reactor overflow line  

DOE Patents (OSTI)

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

Severson, Wayne J. (Pittsburgh, PA)

1976-01-01T23:59:59.000Z

240

Reactor vessel support system  

DOE Patents (OSTI)

A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

1982-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor aging management" 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

CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor  

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

Reactor Contractor ORR Reactor Contractor ORR CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Maintenance Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR More Documents & Publications CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor

242

Spinning fluids reactor  

SciTech Connect

A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

Miller, Jan D; Hupka, Jan; Aranowski, Robert

2012-11-20T23:59:59.000Z

243

Fission reactors and materials  

SciTech Connect

The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions.

Frost, B.R.T.

1981-12-01T23:59:59.000Z

244

Determining Reactor Neutrino Flux  

E-Print Network (OSTI)

Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understanding of the correlation of uncertainties is required for $\\theta_{13}$ experiments. Precise determination of reactor neutrino flux will also improve the sensitivity of the non-proliferation monitoring and future reactor experiments. We will discuss the flux calculation and recent progresses.

Cao, Jun

2011-01-01T23:59:59.000Z

245

Reactor water cleanup system  

DOE Patents (OSTI)

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)

1994-01-01T23:59:59.000Z

246

Determining Reactor Neutrino Flux  

E-Print Network (OSTI)

Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understanding of the correlation of uncertainties is required for $\\theta_{13}$ experiments. Precise determination of reactor neutrino flux will also improve the sensitivity of the non-proliferation monitoring and future reactor experiments. We will discuss the flux calculation and recent progresses.

Jun Cao

2011-01-12T23:59:59.000Z

247

Human Factors Aspects of Operating Small Reactors  

DOE Green Energy (OSTI)

The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. They are considering small modular reactors (SMRs) as one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants, and so may require a concept of operations (ConOps) that also is different. The U.S. Nuclear Regulatory Commission (NRC) has begun examining the human factors engineering- (HFE) and ConOps- aspects of SMRs; if needed, they will formulate guidance to support SMR licensing reviews. We developed a ConOps model, consisting of the following dimensions: Plant mission; roles and responsibilities of all agents; staffing, qualifications, and training; management of normal operations; management of off-normal conditions and emergencies; and, management of maintenance and modifications. We are reviewing information on SMR design to obtain data about each of these dimensions, and have identified several preliminary issues. In addition, we are obtaining operations-related information from other types of multi-module systems, such as refineries, to identify lessons learned from their experience. Here, we describe the project's methodology and our preliminary findings.

OHara, J.M.; Higgins, J.; Deem, R. (BNL); Xing, J.; DAgostino, A. (NRC)

2010-11-07T23:59:59.000Z

248

DOE - Office of Legacy Management -- Hallam  

Office of Legacy Management (LM)

Nebraska Nebraska Hallam, Nebraska, Decommissioned Reactor Site A D&D Program Site hallam_map As part of the DOE Defense Decontamination and Decommissioning (D&D) Program, the Office of Legacy Management manages the Hallam Decommissioned Reactor Site and ensures compliance with applicable federal, state, and local environmental protection laws and regulations, executive orders, and internal DOE policies. The site transferred to the Office of Legacy Management in 2003 and requires routine inspection and maintenance, records-related activities, and stakeholder support. For more information about the Hallam site, view the fact sheet. Site Documents and Links Contact Us Inspection/Sampling Schedule Hallam Decommissioned Reactor Site Mapping and Monitoring (GEMS)

249

Life of Plant Activity Estimates for a Nominal 1000 MWe Pressurized Water Reactor and Boiling Water Reactor  

Science Conference Proceedings (OSTI)

Decommissioning nuclear power plant and disposal site managers must understand the radioactive source term of a nuclear power plant to effectively manage disposition of these materials. This study estimates the radioactive source term from nominal 1000 MWe pressurized water and boiling water reactors to support decisions related to radioactive waste storage, processing, and disposal through decommissioning.BackgroundThis study examines the radionuclide ...

2012-12-05T23:59:59.000Z

250

Advanced reactor design study. Assessing nonbackfittable concepts for improving uranium utilization in light water reactors  

Science Conference Proceedings (OSTI)

The objective of the Advanced Reactor Design Study (ARDS) is to identify and evaluate nonbackfittable concepts for improving uranium utilization in light water reactors (LWRs). The results of this study provide a basis for selecting and demonstrating specific nonbackfittable concepts that have good potential for implementation. Lead responsibility for managing the study was assigned to the Pacific Northwest Laboratory (PNL). Nonbackfittable concepts for improving uranium utilization in LWRs on the once-through fuel cycle were selected separately for PWRs and BWRs due to basic differences in the way specific concepts apply to those plants. Nonbackfittable concepts are those that are too costly to incorporate in existing plants, and thus, could only be economically incorporated in new reactor designs or plants in very early stages of construction. Essential results of the Advanced Reactor Design Study are summarized.

Fleischman, R.M.; Goldsmith, S.; Newman, D.F.; Trapp, T.J.; Spinrad, B.I.

1981-09-01T23:59:59.000Z

251

An Assessment of Integrated Health Management Frameworks  

Science Conference Proceedings (OSTI)

In order to meet the ever increasing demand for energy, the United States nuclear industry is turning to life extension of existing nuclear power plants (NPPs). Economically ensuring the safe, secure, and reliable operation of aging NPPs presents many challenges. The 2009 Light Water Reactor Sustainability Workshop identified online monitoring of active and structural components as essential to better understanding and management of the challenges posed by aging NPPs. Additionally, there is increasing adoption of condition-based maintenance (CBM) for active components in NPPs. These techniques provide a foundation upon which a variety of advanced online surveillance, diagnostic, and prognostic techniques can be deployed to continuously monitor and assess the health of NPP systems and components. The next step in the development of advanced online monitoring is to move beyond CBM to estimating the remaining useful life of active components using prognostic tools. Deployment of prognostic health management (PHM) on the scale of an NPP requires the use of an integrated health management (IHM) framework - a software product (or suite of products) used to manage the necessary elements needed for a complete implementation of online monitoring and prognostics. This paper provides a thoughtful look at the desirable functions and features of IHM architectures. A full PHM system involves several modules, including data acquisition, system modeling, fault detection, fault diagnostics, system prognostics, and advisory generation (operations and maintenance planning). The standards applicable to PHM applications are indentified and summarized. A list of evaluation criteria for PHM software products, developed to ensure scalability of the toolset to an environment with the complexity of an NPP, is presented. Fourteen commercially available PHM software products are identified and classified into four groups: research tools, PHM system development tools, deployable architectures, and peripheral tools.

Lybeck, Nancy; Coble, Jamie B.; Tawfik, Magdy; Bond, Leonard J.

2012-05-18T23:59:59.000Z

252

Aging in American Convents  

E-Print Network (OSTI)

Schuster. Snowdon, David 2001 Aging with Grace: What the Nunreligion, devotion, and aging. CSW JAN09 update tocAging in American Convents FIELDWORK REPORT by Anna I.

Corwin, Anna I.

2009-01-01T23:59:59.000Z

253

HORIZONTAL BOILING REACTOR SYSTEM  

DOE Patents (OSTI)

Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

Treshow, M.

1958-11-18T23:59:59.000Z

254

Generation -IV Reactor Concepts  

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

Generation-IV Reactor Concepts Generation-IV Reactor Concepts Thomas H. Fanning Argonne National Laboratory 9700 South Cass Avenue Argonne, Illinois 60439, USA The Generation-IV International Forum (GIF) is a multi-national research and development (R&D) collaboration. The GIF pursues the development of advanced, next generation reactor technology with goals to improve: a) sustainability (effective fuel utilization and minimization of waste) b) economics (competitiveness with respect to other energy sources) c) safety and reliability (e.g., no need for offsite emergency response), and d) proliferation resistance and physical protection The GIF Technology Roadmap exercise selected six generic systems for further study: the Gas- cooled Fast Reactor (GFR), the Lead-cooled Fast Reactor (LFR), the Molten Salt Reactor (MSR),

255

NUCLEAR REACTORS AND EARTHQUAKES  

SciTech Connect

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

1961-01-01T23:59:59.000Z

256

THERMAL NEUTRONIC REACTOR  

DOE Patents (OSTI)

A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

Spinrad, B.I.

1960-01-12T23:59:59.000Z

257

Improved vortex reactor system  

DOE Patents (OSTI)

An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

Diebold, J.P.; Scahill, J.W.

1995-05-09T23:59:59.000Z

258

Natural Versus Artificial Aging of Nuclear Power Plant Components  

Science Conference Proceedings (OSTI)

Examination of test specimens, placed in eight operating reactors during this study, will provide information on the long-term aging effects of plant environments on cabling and electrical components. Material property data, when compared with measured degradation in artificially aged specimens, will help improve predictions of the safe, useful life of plant equipment.

1986-12-31T23:59:59.000Z

259

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

260

Pressurized fluidized bed reactor  

DOE Patents (OSTI)

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

Isaksson, Juhani (Karhula, FI)

1996-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor aging management" 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

Pressurized fluidized bed reactor  

DOE Patents (OSTI)

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

Isaksson, J.

1996-03-19T23:59:59.000Z

262

Tokamak reactor first wall  

DOE Patents (OSTI)

This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.

1984-11-20T23:59:59.000Z

263

HOMOGENEOUS NUCLEAR POWER REACTOR  

DOE Patents (OSTI)

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

King, L.D.P.

1959-09-01T23:59:59.000Z

264

Materials Development and Degradation Management for Nuclear ...  

Science Conference Proceedings (OSTI)

Materials Development and Degradation Management for Nuclear ... of zirconium alloys in reactor can lead to the presence of hydride precipitates, and can ... for the U.S. Department ofEnergy's National Nuclear Security Administration under ...

265

Integral Fast Reactor: A future source of nuclear energy  

SciTech Connect

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

Southon, R.

1993-09-01T23:59:59.000Z

266

Management and Administration | National Nuclear Security Administration  

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

Management and Administration | National Nuclear Security Administration Management and Administration | 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 Management and Administration Home > About Us > Our Programs > Powering the Nuclear Navy > Management and Administration Management and Administration NNSA's Naval Reactors is committed to excellence and dedicated to meeting

267

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

268

Production reactor characteristics  

SciTech Connect

Reactors for the production of special nuclear materials share many similarities with commercial nuclear power plants. Each relies on nuclear fission, uses uranium fuel, and produces large quantities of thermal power. However, there are some important differences in production reactor characteristics that may best be discussed in terms of mission, role, and technology.

Thiessen, C.W.; Hootman, H.E.

1990-01-01T23:59:59.000Z

269

Advanced converter reactors  

SciTech Connect

Advanced converter reactors (ACRs) of primary US interest are those which can be commercialized within about 20 years, and are: Advanced Light-Water Reactors, Spectral-Shift-Control Reactors, Heavy-Water Reactors (CANDU type), and High-Temperature Gas-Cooled Reactors. These reactors can operate on uranium, thorium, or uranium-thorium fuel cycles, but have the greatest fuel utilization on thorium type cycles. The water reactors tend to operate more economically on uranium cycles, while the HTGR is more economical on thorium cycles. Thus, the HTGR had the greatest practical potential for improving fuel utilization. If the US has 3.4 to 4 million tons U/sub 3/O/sub 8/ at reasonable costs, ACRs can make important contributions to maintaining a high nuclear power level for many decades; further, they work well with fast breeder reactors in the long term under symbiotic fueling conditions. Primary nuclear data needs of ACRs are integral measurements of reactivity coefficients and resonance absorption integrals.

Kasten, P.R.

1979-01-01T23:59:59.000Z

270

NEUTRONIC REACTOR SYSTEM  

DOE Patents (OSTI)

A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

Treshow, M.

1959-02-10T23:59:59.000Z

271

NEUTRONIC REACTOR BURIAL ASSEMBLY  

DOE Patents (OSTI)

A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.

Treshow, M.

1961-05-01T23:59:59.000Z

272

The Integral Fast Reactor  

SciTech Connect

Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs.

Till, C.E.; Chang, Y.I. (Argonne National Lab., IL (USA)); Lineberry, M.J. (Argonne National Lab., Idaho Falls, ID (USA))

1990-01-01T23:59:59.000Z

273

CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor  

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

Reactor Contractor ORR Reactor Contractor ORR CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Training Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR More Documents & Publications CRAD, Conduct of Operations - Oak Ridge National Laboratory High Flux

274

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk  

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

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment September 19, 2012 Presenter: Bentley Harwood, Advanced Test Reactor Nuclear Safety Engineer Battelle Energy Alliance Idaho National Laboratory Topics covered: PRA studies began in the late 1980s 1989, ATR PRA published as a summary report 1991, ATR PRA full report 1994 and 2004 various model changes 2011, Consolidation, update and improvement of previous PRA work 2012/2013, PRA risk monitor implementation Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment More Documents & Publications DOE's Approach to Nuclear Facility Safety Analysis and Management Nuclear Regulatory Commission Handling of Beyond Design Basis Events for

275

CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor  

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

Engineering - Oak Ridge National Laboratory High Flux Isotope Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Engineering Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor

276

DOE - Office of Legacy Management -- Fernald Environmental Management  

Office of Legacy Management (LM)

Fernald Environmental Management Fernald Environmental Management Project - 027 FUSRAP Considered Sites Site: Fernald Environmental Management Project (027) Designated Name: Alternate Name: Location: Evaluation Year: Site Operations: Site Disposition: Radioactive Materials Handled: Primary Radioactive Materials Handled: Radiological Survey(s): Site Status: The Fernald Environmental Management Project (FEMP), formerly known as the Feed Materials Production Center, is located about 18 miles northwest of Cincinnati, Ohio. The facility produced uranium metal products between 1953 and 1989 for use in production reactors to make plutonium and tritium at other Department of Energy (DOE) sites in support of United States defense programs. Uranium metal production ended in July 1989 and resources shifted

277

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

278

Light Water Reactors [Corrosion and Mechanics of Materials] - Nuclear  

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

Light Water Reactors Light Water Reactors Capabilities Materials Testing Environmentally Assisted Cracking (EAC) of Reactor Materials Corrosion Performance/Metal Dusting Overview Light Water Reactors Fatigue Testing of Carbon Steels and Low-Alloy Steels Environmentally Assisted Cracking of Ni-Base Alloys Irradiation-Induced Stress Corrosion Cracking of Austenitic Stainless Steels Steam Generator Tube Integrity Program Air Oxidation Kinetics for Zr-based Alloys Fossil Energy Fusion Energy Metal Dusting Publications List Irradiated Materials Steam Generator Tube Integrity Other Facilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Corrosion and Mechanics of Materials Light Water Reactors Bookmark and Share To continue safe operation of current LWRs, the aging degradation of the

279

Management Plan Management Plan  

E-Print Network (OSTI)

of a coevolving naturalcultural system. Suitable ecosystem attributes can be achieved by managing human maintenance that protects the riparian corridor. · Manage recreational use to protect riparian values Creek and the Warm Springs River during runoff periods. · Road systems and upland management practices

280

Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.  

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 in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage, and cleaning stations-have accumulated satisfactory construction and operation experiences. In addition, two special issues for future development are described in this report: large capacity interim storage and transuranic-bearing fuel handling.

Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

2009-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor aging management" 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

An extended conventional fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor  

SciTech Connect

The B and W mPower{sup TM} reactor is a small pressurized water reactor (PWR) with an integral once-through steam generator and a thermal output of about 500 MW; it is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height PWR assemblies with the familiar 17 x 17 fuel rod array. The Babcock and Wilcox Company (B and W) is offering a core loading and cycle management plan for a four-year cycle based on its presumed attractiveness to potential customers. This option is a once-through fuel cycle in which the entire core is discharged and replaced after four years. In addition, a conventional fuel utilization strategy, employing a periodic partial reload and shuffle, was developed as an alternative to the four-year once-through fuel cycle. This study, which was performed using the Studsvik core design code suite, is a typical multi-cycle projection analysis of the type performed by most fuel management organizations such as fuel vendors and utilities. In the industry, the results of such projections are used by the financial arms of these organizations to assist in making long-term decisions. In the case of the B and W mPower reactor, this analysis demonstrates flexibility for customers who consider the once-through fuel cycle unacceptable from a fuel utilization standpoint. As expected, when compared to the once-through concept, reloads of the B and W mPower reactor will achieve higher batch average discharge exposure, will have adequate shut-down margin, and will have a relatively flat hot excess reactivity trend at the expense of slightly increased peaking. (authors)

Scarangella, M. J. [Babcock and Wilcox Company, 109 Ramsey Place, Lynchburg, VA 24502 (United States)

2012-07-01T23:59:59.000Z

282

REACTOR DEVELOPMENT QUARTERLY PROGRESS REPORT  

DOE Green Energy (OSTI)

The design and construction program of the Bolling Experimental Reactor is reviewed. A number of preliminary experinnents were performed with Borax-II at pressures between 75 and 300 psi. The most corrosion-resistant U-Zr--Nb alloy developed so far is produced by heating in a vacuum to the gamma phase, quenching, and aging for 2 hr at 400 deg C. Special attention is given to the removal of H/sub 2/ from the material. Unclad and unirradiated samples of U--Nb and U--Nb--Zr alloys were corrosion tested in H/sub 2/O. Corrosion rates were also measured under irradiation conditions in CP-5. Elongation measurements of irradiated wrought and cast U--Zr material suggested no way for treating the wrought fuel so that stability comparable to the cast material could be obtained. Natural circulation boiling density tests at 600 psia were made in order to determine the effects of channel cross section and subcooling on the steam void fraction. Results of autoclave and dynamic corrosion studies of 2-S Al in H/sub 2/O are reported. These results include the testing of Ni-clad samples. A large number of criticality calculations were performed for the EBR-Il and the PBR. The solubility of ThO/sub 2/ pellets containing various concentrations of U/sub 3/ O/sub 8/ was tested in water at 316 deg C for periods of 672 to 744 hr. None of the samples disintegrated, although at least one sample developed cracks. Solutions of reactor kinetic equations were attempted for the purpose of studying transients in reactors with lifetimes of 7 x 10/sup -9/ 10/sup -7/, and 6 x 10/ sup -5/ sec. Ignition experiments were performed on Th, Cu, Al, Fe, Mg, Zr, and fluorothene when contacted with fluorides. Except for Zr and fluorothene, the materials did not ignite. (C.H.)

None

1955-01-31T23:59:59.000Z

283

New Production Reactors Program Plan  

SciTech Connect

Part I of this New Production Reactors (NPR) Program Plan: describes the policy basis of the NPR Program; describes the mission and objectives of the NPR Program; identifies the requirements that must be met in order to achieve the mission and objectives; and describes and assesses the technology and siting options that were considered, the Program's preferred strategy, and its rationale. The implementation strategy for the New Production Reactors Program has three functions: Linking the design, construction, operation, and maintenance of facilities to policies requirements, and the process for selecting options. The development of an implementation strategy ensures that activities and procedures are consistent with the rationale and analysis underlying the Program. Organization of the Program. The strategy establishes plans, organizational structure, procedures, a budget, and a schedule for carrying out the Program. By doing so, the strategy ensures the clear assignment of responsibility and accountability. Management and monitoring of the Program. Finally, the strategy provides a basis for monitoring the Program so that technological, cost, and scheduling issues can be addressed when they arise as the Program proceeds. Like the rest of the Program Plan, the Implementation Strategy is a living document and will be periodically revised to reflect both progress made in the Program and adjustments in plans and policies as they are made. 21 figs., 5 tabs.

1990-12-01T23:59:59.000Z

284

Nuclear reactor control column  

DOE Patents (OSTI)

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

Bachovchin, Dennis M. (Plum Borough, PA)

1982-01-01T23:59:59.000Z

285

Slurry reactor design studies  

SciTech Connect

The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

1990-06-01T23:59:59.000Z

286

AGING FACILITY CRITICALITY SAFETY CALCULATIONS  

Science Conference Proceedings (OSTI)

The purpose of this design calculation is to revise and update the previous criticality calculation for the Aging Facility (documented in BSC 2004a). This design calculation will also demonstrate and ensure that the storage and aging operations to be performed in the Aging Facility meet the criticality safety design criteria in the ''Project Design Criteria Document'' (Doraswamy 2004, Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''SNF Aging System Description Document'' (BSC [Bechtel SAIC Company] 2004f, p. 3-12). The scope of this design calculation covers the systems and processes for aging commercial spent nuclear fuel (SNF) and staging Department of Energy (DOE) SNF/High-Level Waste (HLW) prior to its placement in the final waste package (WP) (BSC 2004f, p. 1-1). Aging commercial SNF is a thermal management strategy, while staging DOE SNF/HLW will make loading of WPs more efficient (note that aging DOE SNF/HLW is not needed since these wastes are not expected to exceed the thermal limits form emplacement) (BSC 2004f, p. 1-2). The description of the changes in this revised document is as follows: (1) Include DOE SNF/HLW in addition to commercial SNF per the current ''SNF Aging System Description Document'' (BSC 2004f). (2) Update the evaluation of Category 1 and 2 event sequences for the Aging Facility as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004c, Section 7). (3) Further evaluate the design and criticality controls required for a storage/aging cask, referred to as MGR Site-specific Cask (MSC), to accommodate commercial fuel outside the content specification in the Certificate of Compliance for the existing NRC-certified storage casks. In addition, evaluate the design required for the MSC that will accommodate DOE SNF/HLW. This design calculation will achieve the objective of providing the criticality safety results to support the preliminary design of the Aging Facility. As the ongoing design evolution remains fluid, the results from this design calculation should be evaluated for applicability to any new or modified design. Consequently, the results presented in this document are limited to the current design. The information contained in this document was developed by Environmental and Nuclear Engineering and is intended for the use of Design and Engineering in its work regarding the various criticality related activities performed in the Aging Facility. Yucca Mountain Project personnel from Environmental and Nuclear Engineering should be consulted before the use of the information for purposes other than those stated herein or use by individuals other than authorized personnel in Design and Engineering.

C.E. Sanders

2004-09-10T23:59:59.000Z

287

Nuclear reactor reflector  

DOE Patents (OSTI)

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

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

1994-01-01T23:59:59.000Z

288

Nuclear reactor reflector  

DOE Patents (OSTI)

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

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

1994-06-07T23:59:59.000Z

289

NUCLEAR REACTOR FUEL SYSTEMS  

DOE Patents (OSTI)

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

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

1959-09-15T23:59:59.000Z

290

Spherical torus fusion reactor  

DOE Patents (OSTI)

The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

Martin Peng, Y.K.M.

1985-10-03T23:59:59.000Z

291

CONTROL FOR NEUTRONIC REACTOR  

DOE Patents (OSTI)

S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.

Lichtenberger, H.V.; Cameron, R.A.

1959-03-31T23:59:59.000Z

292

Fast Breeder Reactor studies  

Science Conference Proceedings (OSTI)

This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

1980-07-01T23:59:59.000Z

293

Microfluidic electrochemical reactors  

DOE Patents (OSTI)

A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

Nuzzo, Ralph G. (Champaign, IL); Mitrovski, Svetlana M. (Urbana, IL)

2011-03-22T23:59:59.000Z

294

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

295

CRAD, Configuration Management - Oak Ridge National Laboratory High Flux  

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

Configuration Management - Oak Ridge National Laboratory High Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Configuration Management Program in preparation for restart of the Oak Ridge National Laboratory, High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications

296

Statement on Defense Nuclear Nonproliferation and Naval Reactors Activities  

National Nuclear Security Administration (NNSA)

Defense Nuclear Nonproliferation and Naval Reactors Activities Defense Nuclear Nonproliferation and Naval Reactors Activities before the House Committee on Appropriations Subcommittee on Energy & Water Development | 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 > Congressional Testimony > Statement on Defense Nuclear

297

Statement on Defense Nuclear Nonproliferation and Naval Reactors Activities  

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

Defense Nuclear Nonproliferation and Naval Reactors Activities Defense Nuclear Nonproliferation and Naval Reactors Activities before the House Committee on Appropriations Subcommittee on Energy & Water Development | 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 > Congressional Testimony > Statement on Defense Nuclear

298

Thoughts on Aging  

E-Print Network (OSTI)

The impetus of writing about aging came the other day afterMagnetic Resonance Thoughts on Aging George Feher Received:at Springerlink.com When we, aging scientists, fade away and

Feher, George

2010-01-01T23:59:59.000Z

299

A review of decommissioning considerations for new reactors  

Science Conference Proceedings (OSTI)

At a time of 'nuclear renaissance' when the focus is on advanced reactor designs and construction, it is easy to overlook the decommissioning considerations because such a stage in the life of the new reactors will be some sixty years down the road. Yet, one of the lessons learned from major decommissioning projects has been that decommissioning was not given much thought when these reactors were designed three or four decades ago. Hence, the time to examine what decommissioning considerations should be taken into account is right from the design stage with regular updates of the decommissioning strategy and plans throughout the life cycle of the reactor. Designing D and D into the new reactor designs is necessary to ensure that the tail end costs of the nuclear power are manageable. Such considerations during the design stage will facilitate a more cost-effective, safe and timely decommissioning of the facility when a reactor is eventually retired. This paper examines the current regulatory and industry design guidance for the new reactors with respect to the decommissioning issues and provides a review of the design considerations that can help optimize the reactor designs for the eventual decommissioning. (authors)

Devgun, J.S.Ph.D. [Manager Nuclear Power Technologies, Sargent and Lundy LLC, Chicago, IL (United States)

2008-07-01T23:59:59.000Z

300

Reactor Thermal-Hydraulics  

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

Thermal-Hydraulics Thermal-Hydraulics Dr. Tanju Sofu, Argonne National Laboratory In a power reactor, the energy produced in fission reaction manifests itself as heat to be removed by a coolant and utilized in a thermodynamic energy conversion cycle to produce electricity. A simplified schematic of a typical nuclear power plant is shown in the diagram below. Primary coolant loop Steam Reactor Heat exchanger Primary pump Secondary pump Condenser Turbine Water Although this process is essentially the same as in any other steam plant configuration, the power density in a nuclear reactor core is typically four orders of magnitude higher than a fossil fueled plant and therefore it poses significant heat transfer challenges. Maximum power that can be obtained from a nuclear reactor is often limited by the

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


301

Reactor hot spot analysis  

SciTech Connect

The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.

Vilim, R.B.

1985-08-01T23:59:59.000Z

302

Molten metal reactors  

SciTech Connect

A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

2013-11-05T23:59:59.000Z

303

Compact power reactor  

DOE Patents (OSTI)

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

Wetch, Joseph R. (Woodland Hills, CA); Dieckamp, Herman M. (Canoga Park, CA); Wilson, Lewis A. (Canoga Park, CA)

1978-01-01T23:59:59.000Z

304

NEUTRONIC REACTOR CONSTRUCTION AND OPERATION  

DOE Patents (OSTI)

A method is given for operating a nuclear reactor having a negative coefficient of reactivity to compensate for the change in reactor reactivity due to the burn-up of the xenon peak following start-up of the reactor. When it is desired to start up the reactor within less than 72 hours after shutdown, the temperature of the reactor is lowered prior to start-up, and then gradually raised after start-up.

West, J.M.; Weills, J.T.

1960-03-15T23:59:59.000Z

305

Management Programs  

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

3MS Environment and Waste Management Programs Environment and Waste Management Programs Environment and Waste Management Programs AN L-934 Surveillance of Site A and Plot M...

306

Management Programs  

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

Environment and Waste Management Programs Environment and Waste Management Programs Environment and Waste Management Programs AN L-9213 Surveillance of Site A and Plot M Report...

307

Management Programs  

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

r Environment and Waste Management Programs Environment and Waste Management Programs Environment and Waste Management Programs ANL-949 Surveillance of Site A and Plot M Report...

308

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

309

Applying risk informed methodologies to improve the economics of sodium-cooled fast reactors  

E-Print Network (OSTI)

In order to support the increasing demand for clean sustainable electricity production and for nuclear waste management, the Sodium-Cooled Fast Reactor (SFR) is being developed. The main drawback has been its high capital ...

Nitta, Christopher C

2010-01-01T23:59:59.000Z

310

Thermal-hydraulic analysis of innovative fuel configurations for the sodium fast reactor  

E-Print Network (OSTI)

The sodium fast reactor (SFR) is currently being reconsidered as an instrument for actinide management throughout the world, thanks in part to international programs such as the Generation-IV and especially the Global ...

Memmott, Matthew J

2009-01-01T23:59:59.000Z

311

Analysis of strategies for improving uranium utilization in pressurized water reactors  

E-Print Network (OSTI)

Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal ...

Sefcik, Joseph A.

1981-01-01T23:59:59.000Z

312

N Reactor standby program plan; Part 5: Safety and environment FY-90 update  

Science Conference Proceedings (OSTI)

N Reactor was shutdown by order of DOE in January 1987 to permit implementation of a series of hardware and software safety enhancements. These upgrades were designed to enhance the safe operation of N Reactor and provide comparability to commercial nuclear industry standards and requirements. The enhancements were identified by a series of external independent reviews and were managed under the N Reactor Safety Enhancement Program. In February 1988, DOE directed that N Reactor would be placed in standby. Additional guidance in FY-88 directed that the reactor would be placed in a dry layup status based on preceding Westinghouse Hanford Studies directed at defining a cost effective layup method which would preserve the ability to subsequently restart and operate N Reactor if required. The FY-88 directives also stipulated that the ability to restart N Reactor in a 2--3 year period from date of notification be preserved. 7 figs., 8 tabs.

NONE

1989-12-01T23:59:59.000Z

313

REACTOR GROUT THERMAL PROPERTIES  

DOE Green Energy (OSTI)

Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

2011-01-28T23:59:59.000Z

314

CRAD, Emergency Management - Oak Ridge National Laboratory High Flux  

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

Oak Ridge National Laboratory High Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Emergency Management Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Emergency Management - Oak Ridge National Laboratory High Flux

315

Advanced Reactor Development and Technology - Nuclear Engineering...  

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

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

316

Zero Power Reactor simulation | Argonne National Laboratory  

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

Zero Power Reactor simulation Share Description Ever wanted to see a nuclear reactor core in action? Here's a detailed simulation of the Zero Power Reactor experiment, run by...

317

THE MATERIALS OF FAST BREEDER REACTORS  

E-Print Network (OSTI)

jet aircraft engines, and nuclear reactor fuel elements. Ancomponents of a nuclear reactor core are susceptible tothe nuclear physics of the thermal and fast neutron reactors

Olander, Donald R.

2013-01-01T23:59:59.000Z

318

Gas-cooled reactors  

SciTech Connect

Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing.

Schulten, R.; Trauger, D.B.

1976-01-01T23:59:59.000Z

319

ORNL - Restart of the High Flux Isotope Reactor 2-07  

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

AUTHORIZATION BASIS MANAGEMENT AUTHORIZATION BASIS MANAGEMENT OBJECTIVE AB-1: The nuclear safety program has been appropriately modified to reflect the CS modification and its reactor interface, sufficient numbers of qualified nuclear safety staff and management are provided, and adequate facilities and equipment are available to ensure services are adequate to conduct and support operations with the CS modification. Functions, assignments, responsibilities, and reporting relationships are clearly defined, understood, and effectively implemented with line management control of safety. (CR-1, CR-2, CR-6) Criteria The nuclear safety program and organization are established and functioning to support reactor operations with the CS modification. Functions, responsibilities, and

320

Decommissioning of the Tokamak Fusion Test Reactor  

SciTech Connect

The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

2003-10-28T23:59:59.000Z

Note: This page contains sample records for the topic "reactor aging management" 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

Advanced Light Water Reactor utility requirements document  

SciTech Connect

The ALWR Requirements Document is a primary work product of the EPRI Program. This document is an extensive compilation of the utility requirements for design, construction and performance of advanced light water reactor power plants for the 1990s and beyond. The Requirements Document's primary emphasis is on resolution of significant problems experienced at existing nuclear power plants. It is intended to be used with companion documents, such as utility procurement specifications, which would cover the remaining detailed technical requirements applicable to new plant projects. The ALWR Requirements Document consists of several major parts. This volume is Part I, The Executive Summary. It is intended to serve as a concise, management level synopsis of advanced light water reactors including design objectives and philosophy, overall configuration and features and the steps necessary to proceed from the conceptual design stage to a completed, functioning power plant.

1986-06-01T23:59:59.000Z

322

Light Water Reactor Sustainability Program - Integrated Program Plan |  

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

Light Water Reactor Sustainability Program - Integrated Program Light Water Reactor Sustainability Program - Integrated Program Plan Light Water Reactor Sustainability Program - Integrated Program Plan The Light Water Reactor Sustainability (LWRS) Program is a research and development (R&D) program sponsored by the U. S. Department of Energy (DOE), performed in close collaboration and cooperation with related industry R&D programs. The LWRS Program provides technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants, utilizing the unique capabilities of the national laboratory system. Sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer than-initially-licensed lifetime. It has two facets

323

CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor  

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

Oak Ridge National Laboratory High Flux Isotope Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Engineering Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory, High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR More Documents & Publications

324

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab  

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

UPTON, N.Y. - American Recovery and Reinvestment Act UPTON, N.Y. - American Recovery and Reinvestment Act workers are in the final stage of decommissioning a nuclear reactor after they recently removed thick steel shields once used to absorb neutrons produced for research. The Brookhaven National Laboratory is using $39 million from the Recovery Act to decommission the Brookhaven Graphite Research Reactor, the world's first reactor built solely for peaceful research purposes. The decommissioning is slated for completion later this year and will end Office of Environmental Management legacy cleanup activities at the Lab. The neutron shields were located on the north and south sides of a 700-ton graphite pile. The three-inch-thick shields absorbed neutrons that escaped from the graphite pile. The shields also limited movement of the pile when the reactor was in opera-

325

N reactor standby program plan: Part 2, Transition to standby  

Science Conference Proceedings (OSTI)

The N Reactor Standby Program is the Westinghouse Hanford Company (WHC) program to prepare the N Reactor for cold standby in a cost effective and quality manner with a primary emphasis on safety. The N Reactor Standby Program is comprised of five parts or subprograms, of which the Transition to Standby Program is part two. Transition to Standby is that part of the overall program that encompasses the on-going operational and maintenance activities required to maintain safe control of the reactor and support facilities until their layup is complete. The Transition to Standby Program Plan has been developed to outline the program plans, baselines, and controls to be employed in the management of the Transition to Standby Program. This plan also defines the roles and responsibilities of the major program participants.

Slinn, S.A.

1988-09-01T23:59:59.000Z

326

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database  

SciTech Connect

Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

Wang, Jy-An John [ORNL

2010-08-01T23:59:59.000Z

327

Development of Underwater Laser Cladding and Underwater Laser Seal Welding Techniques for Reactor Components (II)  

SciTech Connect

Stress corrosion cracking (SCC) is one of the major reasons to reduce the reliability of aged reactor components. Toshiba has been developing underwater laser welding onto surface of the aged components as maintenance and repair techniques. Because most of the reactor internal components to apply this underwater laser welding technique have 3-dimensional shape, effect of welding positions and welded shapes are examined and presented in this report. (authors)

Masataka Tamura; Shohei Kawano; Wataru Kouno; Yasushi Kanazawa [Toshiba Corporation (Japan)

2006-07-01T23:59:59.000Z

328

Nuclear reactor safety device  

DOE Patents (OSTI)

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

Hutter, Ernest (Wilmette, IL)

1986-01-01T23:59:59.000Z

329

REACTOR CONTROL DEVICE  

DOE Patents (OSTI)

A wholly mechanical compact control device is designed for automatically rendering the core of a fission reactor subcritical in response to core temperatures in excess of the design operating temperature limit. The control device comprises an expansible bellows interposed between the base of a channel in a reactor core and the inner end of a fuel cylinder therein which is normally resiliently urged inwardly. The bellows contains a working fluid which undergoes a liquid to vapor phase change at a temperature substantially equal to the design temperature limit. Hence, the bellows abruptiy expands at this limiting temperature to force the fuel cylinder outward and render the core subcritical. The control device is particularly applicable to aircraft propulsion reactor service. (AEC)

Graham, R.H.

1962-09-01T23:59:59.000Z

330

A NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor for producing thermoelectric power is described. The reactor core comprises a series of thermoelectric assemblies, each assembly including fissionable fuel as an active element to form a hot junction and a thermocouple. The assemblies are disposed parallel to each other to form spaces and means are included for Introducing an electrically conductive coolant between the assemblies to form cold junctions of the thermocouples. An electromotive force is developed across the entire series of the thermoelectric assemblies due to fission heat generated in the fuel causing a current to flow perpendicular to the flow of coolant and is distributed to a load outside of the reactor by means of bus bars electrically connected to the outermost thermoelectric assembly.

Luebke, E.A.; Vandenberg, L.B.

1959-09-01T23:59:59.000Z

331

Reactor for exothermic reactions  

DOE Patents (OSTI)

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1993-03-02T23:59:59.000Z

332

MERCHANT MARINE SHIP REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

1961-05-01T23:59:59.000Z

333

Heat dissipating nuclear reactor  

DOE Patents (OSTI)

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

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

1987-01-01T23:59:59.000Z

334

Heat dissipating nuclear reactor  

DOE Patents (OSTI)

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

Hunsbedt, A.; Lazarus, J.D.

1985-11-21T23:59:59.000Z

335

MERCHANT MARINE SHIP REACTOR  

DOE Patents (OSTI)

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

Sankovich, M.F.; Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Gestson, D.K.

1961-05-01T23:59:59.000Z

336

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from

337

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations, which govern the operation of commercial nuclear power plants, require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including

338

Nuclear reactor apparatus  

DOE Patents (OSTI)

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

Wade, Elman E. (Ruffs Dale, PA)

1978-01-01T23:59:59.000Z

339

Fusion reactor pumped laser  

DOE Patents (OSTI)

A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

Jassby, Daniel L. (Princeton, NJ)

1988-01-01T23:59:59.000Z

340

Fast quench reactor method  

DOE Patents (OSTI)

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID); Berry, Ray A. (Idaho Falls, ID)

1999-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor aging management" 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

Perspectives on reactor safety  

SciTech Connect

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

1994-03-01T23:59:59.000Z

342

Reactor Neutrino Experiments  

E-Print Network (OSTI)

Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measurement are also briefed.

Jun Cao

2007-12-06T23:59:59.000Z

343

Diagnostics for hybrid reactors  

SciTech Connect

The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

Orsitto, Francesco Paolo [ENEA Unita' Tecnica Fusione , Associazione ENEA-EURATOM sulla Fusione C R Frascati v E Fermi 45 00044 Frascati (Italy)

2012-06-19T23:59:59.000Z

344

THERMAL NUCLEAR REACTOR  

DOE Patents (OSTI)

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

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

1957-09-24T23:59:59.000Z

345

MEANS FOR SHIELDING REACTORS  

DOE Patents (OSTI)

A reactor of the heterageneous, heavy water moderated type is described. The reactor is comprised of a plurality of vertically disposed fuel element tubes extending through a tank of heavy water moderator and adapted to accommodate a flow of coolant water in contact with the fuel elements. A tank containing outgoing coolant water is disposed above the core to function is a radiation shield. Unsaturated liquid hydrocarbon is floated on top of the water in the shield tank to reduce to a minimum the possibility of the occurrence of explosive gaseous mixtures resulting from the neutron bombardment of the water in the shield tank.

Garrison, W.M.; McClinton, L.T.; Burton, M.

1959-03-10T23:59:59.000Z

346

Export possibilities for small nuclear reactors  

Science Conference Proceedings (OSTI)

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

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

2007-07-01T23:59:59.000Z

347

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

348

THE HOMOGENEOUS SUSPENSION REACTOR PROJECT  

SciTech Connect

The considerations which led to the study of a homogeneous suspension reactor are reviewed briefly. The characteristics of the KEMA Suspension Test Reactor (KSTR) are then summarized. (J.S.R.)

Went, J.J.

1963-02-01T23:59:59.000Z

349

Advanced Nuclear Reactors | Department of Energy  

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

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

350

Innovative design of uranium startup fast reactors  

E-Print Network (OSTI)

Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

Fei, Tingzhou

2012-01-01T23:59:59.000Z

351

ORNL - Restart of the High Flux Isotope Reactor 2-07  

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

QUALITY ASSURANCE (QA) QUALITY ASSURANCE (QA) OBJECTIVE QA-1: The RRD QA program has been appropriately modified to reflect the CS modification and its reactor interface, and sufficient numbers of qualified QA personnel are provided to ensure services are adequate to support reactor operation. The QA functions, assignments, responsibilities, and reporting relationships are clearly defined, understood, and effectively implemented with line management control of safety. QA personnel exhibit awareness of the applicable requirements pertaining to reactor operation with the CS and the associated hazards. Through their actions, they have demonstrated a high-priority commitment to comply with these requirements. The level of knowledge of QA personnel related to reactor

352

Reactor operation safety information document  

Science Conference Proceedings (OSTI)

The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

Not Available

1990-01-01T23:59:59.000Z

353

An Assessment of Integrated Health Management (IHM) Frameworks  

Science Conference Proceedings (OSTI)

In order to meet the ever increasing demand for energy, the United States nuclear industry is turning to life extension of existing nuclear power plants (NPPs). Economically ensuring the safe, secure, and reliable operation of aging nuclear power plants presents many challenges. The 2009 Light Water Reactor Sustainability Workshop identified online monitoring of active and structural components as essential to the better understanding and management of the challenges posed by aging nuclear power plants. Additionally, there is increasing adoption of condition-based maintenance (CBM) for active components in NPPs. These techniques provide a foundation upon which a variety of advanced online surveillance, diagnostic, and prognostic techniques can be deployed to continuously monitor and assess the health of NPP systems and components. The next step in the development of advanced online monitoring is to move beyond CBM to estimating the remaining useful life of active components using prognostic tools. Deployment of prognostic health management (PHM) on the scale of a NPP requires the use of an integrated health management (IHM) framework - a software product (or suite of products) used to manage the necessary elements needed for a complete implementation of online monitoring and prognostics. This paper provides a thoughtful look at the desirable functions and features of IHM architectures. A full PHM system involves several modules, including data acquisition, system modeling, fault detection, fault diagnostics, system prognostics, and advisory generation (operations and maintenance planning). The standards applicable to PHM applications are indentified and summarized. A list of evaluation criteria for PHM software products, developed to ensure scalability of the toolset to an environment with the complexity of a NPP, is presented. Fourteen commercially available PHM software products are identified and classified into four groups: research tools, PHM system development tools, deployable architectures, and peripheral tools.

N. Lybeck; M. Tawfik; L. Bond; J. Coble

2012-05-01T23:59:59.000Z

354

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A reactor fuel element of the capillary tube type is described. The element consists of a thin walled tube, sealed at both ends, and having an interior coatlng of a fissionable material, such as uranium enriched in U-235. The tube wall is gas tight and is constructed of titanium, zirconium, or molybdenum.

Kesselring, K.A.; Seybolt, A.U.

1958-12-01T23:59:59.000Z

355

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A fuel slug for a reactor which acts as a safety device is described. The fuel slug is an aluminum tube with a foil lining the inside surface of the tube, the foil being fabricated of uranium in a lead matrix.

Horning, W.A.; Lanning, D.D.; Donahue, D.J.

1959-10-01T23:59:59.000Z

356

Nuclear reactor building  

DOE Patents (OSTI)

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

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

1994-04-05T23:59:59.000Z

357

Neutronic reactor thermal shield  

DOE Patents (OSTI)

1. The method of operating a water-cooled neutronic reactor having a graphite moderator which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40-60 volume percent of the mixture, in contact with the graphite moderator.

Wende, Charles W. J. (West Chester, PA)

1976-06-15T23:59:59.000Z

358

NEUTRONIC REACTOR FUEL PUMP  

DOE Patents (OSTI)

A reactor fuel pump is described which offers long life, low susceptibility to radiation damage, and gaseous fission product removal. An inert-gas lubricated bearing supports a journal on one end of the drive shsft. The other end has an impeller and expansion chamber which effect pumping and gas- liquid separation. (T.R.H.)

Cobb, W.G.

1959-06-01T23:59:59.000Z

359

The First Reactor  

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

Chicago Pile-1 (CP-1) was the world's first nuclear reactor. CP-1 was built on a rackets court, under the abandoned west stands of the original Alonzo Stagg Field stadium, at the University of...

360

Thermal Reactor Safety  

Science Conference Proceedings (OSTI)

Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

Not Available

1980-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor aging management" 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

WATER BOILER REACTOR  

DOE Patents (OSTI)

As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

King, L.D.P.

1960-11-22T23:59:59.000Z

362

Cermet fuel reactors  

Science Conference Proceedings (OSTI)

Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

1987-09-01T23:59:59.000Z

363

Fossil fuel furnace reactor  

DOE Patents (OSTI)

A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

Parkinson, William J. (Los Alamos, NM)

1987-01-01T23:59:59.000Z

364

Nuclear Reactors and Technology  

SciTech Connect

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

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

1992-01-01T23:59:59.000Z

365

Nuclear reactor building  

DOE Patents (OSTI)

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

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

1994-01-01T23:59:59.000Z

366

Reactor component automatic grapple  

DOE Patents (OSTI)

A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment.

Greenaway, Paul R. (Bethel Park, PA)

1982-01-01T23:59:59.000Z

367

NUCLEAR REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

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

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

1963-06-11T23:59:59.000Z

368

NEUTRONIC REACTOR STRUCTURE  

DOE Patents (OSTI)

A neutronic reactor is described. It has a core consisting of natural uranium and heavy water and having a K-factor greater than unity which is surrounded by a reflector consisting of natural uranium and ordinary water having a Kfactor less than unity.

Weinberg, A.M.; Vernon, H.C.

1961-05-30T23:59:59.000Z

369

NEUTRONIC REACTOR STRUCTURE  

DOE Patents (OSTI)

The neutronic reactor is comprised of a core consisting of natural uranium and heavy water with a K-factor greater than unity. The core is surrounded by a reflector consisting of natural uranium and ordinary water with a Kfactor less than unity. (AEC)

Vernon, H.C.; Weinberg, A.M.

1961-05-30T23:59:59.000Z

370

NEUTRONIC REACTOR SHIELD  

DOE Patents (OSTI)

The reactor radiation shield material is comprised of alternate layers of iron-containing material and compressed cellulosic material, such as masonite. The shielding material may be prefabricated in the form of blocks, which can be stacked together in ary desired fashion to form an effective shield.

Fermi, E.; Zinn, W.H.

1957-09-24T23:59:59.000Z

371

NEUTRONIC REACTOR CONTROL ELEMENT  

DOE Patents (OSTI)

A boron-10 containing reactor control element wherein the boron-10 is dispersed in a matrix material is describeri. The concentration of boron-10 in the matrix varies transversely across the element from a minimum at the surface to a maximum at the center of the element, prior to exposure to neutrons. (AEC)

Beaver, R.J.; Leitten, C.F. Jr.

1962-04-17T23:59:59.000Z

372

Solutioning and Aging Behaviours  

Science Conference Proceedings (OSTI)

Mar 13, 2012 ... Aluminum Alloys: Fabrication, Characterization and Applications: Solutioning and Aging Behaviours Sponsored by: The Minerals, Metals and ...

373

Solutioning and Aging Behaviors  

Science Conference Proceedings (OSTI)

Aluminum Alloys: Development, Characterization and Applications: Solutioning and Aging Behaviors Sponsored by: TMS Light Metals Division, TMS: Aluminum

374

REACTOR FUEL ELEMENTS TESTING CONTAINER  

DOE Patents (OSTI)

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Whitham, G.K.; Smith, R.R.

1963-01-15T23:59:59.000Z

375

EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

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

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

1963-12-24T23:59:59.000Z

376

Aging of SRC liquids  

Science Conference Proceedings (OSTI)

The viscosity of SRC?LL liquid increases when subjected to accelerated aging by bubbling oxygen in the presence of copper strip at 62°C. Precipitates are formed and can be separated from the aged liquid by Soxhlet extraction with pentane. A 30?70 blend of SRC?I with SRC?LL was subjected to oxygen aging in the absence of copper

T. Hara; L. Jones; K. C. Tewari; N. C. Li

1981-01-01T23:59:59.000Z

377

Modeling Chemical Reactors I: Quiescent Reactors  

E-Print Network (OSTI)

We introduce a fully generalized quiescent chemical reactor system in arbitrary space $\\vdim =1,2$ or 3, with $n\\in\\mathbb{N}$ chemical constituents $\\alpha_{i}$, where the character of the numerical solution is strongly determined by the relative scaling between the local reactivity of species $\\alpha_{i}$ and the local functional diffusivity $\\mathscr{D}_{ij}(\\alpha)$ of the reaction mixture. We develop an operator time-splitting predictor multi-corrector RK--LDG scheme, and utilize $hp$-adaptivity relying only on the entropy $\\mathscr{S}_{\\mathfrak{R}}$ of the reactive system $\\mathfrak{R}$. This condition preserves these bounded nonlinear entropy functionals as a necessarily enforced stability condition on the coupled system. We apply this scheme to a number of application problems in chemical kinetics; including a difficult classical problem arising in nonequilibrium thermodynamics known as the Belousov-Zhabotinskii reaction where we utilize a concentration-dependent diffusivity tensor $\\mathscr{D}_{ij}(...

Michoski, C E; Schmitz, P G

2010-01-01T23:59:59.000Z

378

Survey of Optimization of Reactor Coolant Cleanup Systems: For Boiling Water Reactors and Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Optimization of the reactor coolant cleanup systems in the boiling water reactor (BWR) and pressurized water reactor (PWR) environment is important for controlling the transport of corrosion products (metals and activated metals), fission products, and coolant impurities (soluble and insoluble) throughout the reactor coolant loop, and this optimization contributes to reducing primary system radiation fields. The removal of radionuclides and corrosion products is just one of many functions (both ...

2013-09-27T23:59:59.000Z

379

ORNL - Restart of the High Flux Isotope Reactor 2-07  

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

OPERATIONS OPERATIONS OBJECTIVE OP-1: Operations staff and management exhibit awareness of applicable requirements pertaining to CS operation, hazards, and reactor operations with the hydrogen-moderated CS. Through their actions, they have demonstrated a high-priority commitment to comply with these requirements. The level of knowledge of reactor operations and CS system operations managers and staff related to CS operations, hazards, and reactor operations with the hydrogen-moderated CS is adequate based on interviews. Sufficient numbers of qualified reactor operations and CS system operations staff and management are available to conduct and support safe operations with the hydrogen-moderated CS. (CR - 1, CR - 4, CR - 6) Criteria * Minimum staffing requirements have been established for operations and support

380

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

Note: This page contains sample records for the topic "reactor aging management" 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

Reactor vessel support system. [LMFBR  

DOE Patents (OSTI)

A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

Golden, M.P.; Holley, J.C.

1980-05-09T23:59:59.000Z

382

CRAD, Emergency Management - Oak Ridge National Laboratory High Flux  

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

Oak Ridge National Laboratory High Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Emergency Management Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

383

ORNL - Restart of the High Flux Isotope Reactor 2-07  

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

Management and Organization Management and Organization OBJECTIVE MG-1: Line management has integrated within its existing ISM system and implementing mechanisms, programs that appropriately address CS operations, hazards, and reactor interface to assure safe accomplishment of work. Safety management programs of particular interest include the following (CR - 1): * maintenance and testing (addressed by MT-1) * conduct of operations (addressed by OP-1and -5) * training/qualification (addressed by TR-1) * nuclear safety (addressed by AB-2) * emergency management (addressed by EP-1and -2) * configuration management (addressed by ES-3) * fire protection (addressed by ESH-4) * industrial safety and hygiene (addressed by ESH-2) * quality assurance (addressed by ESH-6)

384

ORNL - Restart of the High Flux Isotope Reactor 2-07  

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

Management Management OBJECTIVE MG-1: Line management has established programs to ensure safe accomplishment of work. Personnel exhibit awareness of public and worker safety, health, and environmental protection requirements, and through their actions, they demonstrate a high-priority commitment to comply with these requirements. (Core Requirements 1 and 2) Criteria * Line management has integrated programs within its existing ISMS and implementing mechanisms that appropriately address the major changes implemented during this outage, notably the CS, in order to continue to assure safe accomplishment of work. * Senior management and RRD management exhibit awareness of the applicable requirements pertaining to reactor operation, with emphasis on the

385

DOE - Office of Legacy Management -- Piqua  

Office of Legacy Management (LM)

Ohio Ohio Piqua, Ohio, Decommissioned Reactor Site A D&D Program Site piqua_map As part of the DOE Defense Decontamination and Decommissioning (D&D) Program, the Office of Legacy Management manages the Piqua Decommissioned Reactor Site and ensures compliance with applicable federal, state, and local environmental protection laws and regulations, executive orders, and internal DOE policies. The site transferred to the Office of Legacy Management in 2003 and requires routine inspection and maintenance, records-related activities, and stakeholder support. For more information about the Piqua site, view the fact sheet. Site Documents and Links Contact Us Inspection/Sampling Schedule Piqua Decommissioned Reactor Site Mapping and Monitoring (GEMS) Last Updated: 6/13/2011

386

DOE - Office of Legacy Management -- Bonus  

Office of Legacy Management (LM)

Puerto Rico Puerto Rico Boiling Nuclear Superheater (BONUS), Puerto Rico, Decommissioned Reactor Site La Antiqua Central Nuclear de Aqua Hirviente Sobrecalentada Reactor Fuera de Servico Activo A D&D Program Site bonus_map3 As part of the DOE Defense Decontamination and Decommissioning (D&D) Program, the Office of Legacy Management manages the Boiling Nuclear Superheater (BONUS) Decommissioned Reactor Site and ensures compliance with applicable federal, state, and local environmental protection laws and regulations, executive orders, and internal DOE policies. The site transferred to the Office of Legacy Management in 2004 and requires routine inspection and maintenance, records-related activities, and stakeholder support. For more information about the BONUS site, view the fact sheet.

387

Nuclear reactor construction with bottom supported reactor vessel  

DOE Patents (OSTI)

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

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

1987-01-01T23:59:59.000Z

388

Performance Management  

Science Conference Proceedings (OSTI)

*. Bookmark and Share. Performance Management. ... If you are unable to access these PDF files, please contact the Office of Workforce Management.

2013-04-20T23:59:59.000Z

389

Determining age of whales  

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

Determining age of whales Determining age of whales Name: Bruce W Walkey Age: N/A Location: N/A Country: N/A Date: N/A Question: While browsing through the Internet, I came upon a question by two fifth grade students. Their question got me thinking and now I pose it to you. How can you determine the age of whales? Since they are mammals, can the methods that are used on humans be used on whales? What are some tests that can be done on bones or tissues to determine age? Looking forward to your reply. Replies: Although it is difficult to determine the age of whales (unless they are born in captivity and we know their birth date), several methods have been commonly used: 1) (if female) the examination of the ovaries 2) Examination of the ridges on baleen, which are not uniform in size and analogous to tree rings. The problem with this is that baleen wears away over time. 3) Studying layers of ossification in an ear bone is probably the most accurate method of aging, since internal bones don't wear away. The biggest problem with aging methods is that they usually require that you are dissecting the animal, and often, we would like a method of aging for live active animals. The best we can do here is to compare the size and markings of whales of known age to those found in the wild. Great question!

390

HOMOGENEOUS NUCLEAR REACTOR  

SciTech Connect

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

1960-07-11T23:59:59.000Z

391

AIR COOLED NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

Fermi, E.; Szilard, L.

1958-05-27T23:59:59.000Z

392

Nuclear reactor safety device  

DOE Patents (OSTI)

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

Hutter, E.

1983-08-15T23:59:59.000Z

393

Spherical torus fusion reactor  

DOE Patents (OSTI)

A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

Peng, Yueng-Kay M. (Oak Ridge, TN)

1989-01-01T23:59:59.000Z

394

Fusion reactor pumped laser  

DOE Patents (OSTI)

A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

Jassby, D.L.

1987-09-04T23:59:59.000Z

395

HOMOGENEOUS NUCLEAR REACTOR  

DOE Patents (OSTI)

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

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

1959-02-17T23:59:59.000Z

396

Water reactor fuel cladding  

Science Conference Proceedings (OSTI)

This patent describes a nuclear reactor fuel element cladding tube. It comprises: an outer cylindrical layer of a first zirconium alloy selected from the group consisting of Zircaloy-2 and Zircaloy-4; an inner cylindrical layer of a second zirconium alloy consisting essentially of about 0.19 to 0.6 wt.% tin, about 0.19 to less than 0.5 wt.% iron, about 100 to 700 ppm oxygen, less than 2000 ppm total impurities, and the remainder essentially zirconium; the inner layer characterized by aqueous corrosion resistance substantially the same as the first zirconium alloy; the inner layer characterized by improved resistance to PCI crack propagation under reactor operating conditions compared to the first zirconium alloy and substantially the same PCI crack propagation resistance compared to unalloyed zirconium; and the inner cylindrical layer is metallurgically bonded to the outer layer.

Foster, J.P.; McDonald, S.G.

1990-06-12T23:59:59.000Z

397

International Collaboration with the Shutdown of the BN-350 Reactor  

SciTech Connect

Representatives from the United States and the United Kingdom discussed areas where collaboration on the shutdown of the BN-350 Reactor in Aktau, Kazakhstan would benefit not only Kazakhstan, but would also help to assure the successful shutdown of the reactor. A fundamental understanding of the basis for collaboration has been for each side to ‘add value’ to each of the project areas, rather than simply substitute for each other’s experience. This approach has brought distinct technical and management benefits to the decommissioning activities in Kazakhstan.

J. A. Michelbacher; P.B. Wells; N. Organ; D. Wells

2005-08-01T23:59:59.000Z

398

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

399

Power Reactor Decommissioning Experience  

Science Conference Proceedings (OSTI)

During the past two decades the NRC regulated nuclear industry has encountered and dealt with a diverse range of political, financial and technological challenges while decommissioning its nuclear facilities. During that time, the decommissioning of nuclear facilities has evolved into a mature industry in the United States with a number of large power reactors successfully decommissioned and their NRC licenses terminated. One of the challenges discussed in this report is site release standards, required ...

2011-07-08T23:59:59.000Z

400

Exhaust gas reactor  

SciTech Connect

A reactor for the oxidation of unburned and partially burned components in the exhaust gas from an internal combustion engine comprising a chamber which is substantially circular in cross sections perpendicular to its axis, one or more inlet pipes which pass a mixture of exhaust gas and air substantially tangentially into the chamber near to one end thereof, and an outlet pipe near to the other end of the chamber and which is so arranged that exhaust gas leaves the chamber substantially tangentially. The tangential inlet and tangential outlet of gases minimizes energy losses in the gas passing through the reactor. The ratio of the cross-sectional areas of the inlet pipe(s) to reactor chamber is preferably from 1:9 to 25:36, and similar ranges of crosssectional area ratios are preferred for the outlet pipe and chamber. The ratio of the length of the reaction chamber to diameter is preferably from 1:1 to 4:1. The chamber may be cylindrical or divergent from inlet end to outlet end and may be thermally insulated.

Camarsa, M.; Cocchiara, F.; Garcea, G.P.

1981-11-24T23:59:59.000Z

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


401

BOILER-SUPERHEATED REACTOR  

DOE Patents (OSTI)

A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

Heckman, T.P.

1961-05-01T23:59:59.000Z

402

Modeling Chemical Reactors I: Quiescent Reactors  

E-Print Network (OSTI)

We introduce a fully generalized quiescent chemical reactor system in arbitrary space $\\vdim =1,2$ or 3, with $n\\in\\mathbb{N}$ chemical constituents $\\alpha_{i}$, where the character of the numerical solution is strongly determined by the relative scaling between the local reactivity of species $\\alpha_{i}$ and the local functional diffusivity $\\mathscr{D}_{ij}(\\alpha)$ of the reaction mixture. We develop an operator time-splitting predictor multi-corrector RK--LDG scheme, and utilize $hp$-adaptivity relying only on the entropy $\\mathscr{S}_{\\mathfrak{R}}$ of the reactive system $\\mathfrak{R}$. This condition preserves these bounded nonlinear entropy functionals as a necessarily enforced stability condition on the coupled system. We apply this scheme to a number of application problems in chemical kinetics; including a difficult classical problem arising in nonequilibrium thermodynamics known as the Belousov-Zhabotinskii reaction where we utilize a concentration-dependent diffusivity tensor $\\mathscr{D}_{ij}(\\alpha)$, in addition to solving a simple equilibrium problem in order to evaluate the numerical error behavior.

C. E. Michoski; J. A. Evans; P. G. Schmitz

2010-12-28T23:59:59.000Z

403

Aging Matters What's Inside  

E-Print Network (OSTI)

Aging Matters What's Inside Letter from the dean PSU winS Carter award artiSan eConomy eLi BUi and Public affairs 2008 fall newsletter www.pdx.edu/cupa aGinG is a modern fact of life. In 2006 the world's population of adults aged 60 or over was 650 million. By 2025 that number will nearly double. By 2050

Bertini, Robert L.

404

Materials Degradation and Aging  

Science Conference Proceedings (OSTI)

Aug 28, 2007 ... Improve plant capacity, reliability, and availability. Materials Degradation and Aging - An EPRI Nuclear Power Action Plan, Report No.

405

Design guide for category V reactors transient reactors  

SciTech Connect

The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category V reactor structures, components, and systems.

Brynda, W J; Karol, R C; Lobner, P R; Powell, R W; Straker, E A

1979-03-01T23:59:59.000Z

406

Managing time  

Science Conference Proceedings (OSTI)

Professionals overwhelmed with information glut can find hope from new insights about time management.

Peter J. Denning

2011-03-01T23:59:59.000Z

407

Heavy Water and Graphite Reactors - Reactors designed/built by...  

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

experiments, necessary to achieve higher precision for the determination of reactor power distribution patterns, effect of non-uniform void distributions, kinetic behavior,...

408

Reactor technology: power conversion systems and reactor operation and maintenance  

SciTech Connect

The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He/sup 3/ reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored.

Powell, J.R.

1977-01-01T23:59:59.000Z

409

Melatonin and the aging brain  

E-Print Network (OSTI)

mitochondrial decay of aging. Mol. Aspects Med. 26, Ames,the degenerative diseases of aging. Proc. Natl. Acad. Sci.2004. Retardation of brain aging by chronic treatment with

BONDY, S; SHARMAN, E

2007-01-01T23:59:59.000Z

410

Integrated fuel performance and thermal-hydraulic sub-channel models for analysis of sodium fast reactors  

E-Print Network (OSTI)

Sodium Fast Reactors (SFR) show promise as an effective way to produce clean safe nuclear power while properly managing the fuel cycle. Accurate computer modeling is an important step in the design and eventual licensing ...

Fricano, Joseph William

2012-01-01T23:59:59.000Z

411

Nuclear Reactor Severe Accident Experiments  

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

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

412

Designing Reactors to Facilitate Decommissioning  

SciTech Connect

Critics of nuclear power often cite issues with tail-end-of-the-fuel-cycle activities as reasons to oppose the building of new reactors. In fact, waste disposal and the decommissioning of large nuclear reactors have proven more challenging than anticipated. In the early days of the nuclear power industry the design and operation of various reactor systems was given a great deal of attention. Little effort, however, was expended on end-of-the-cycle activities, such as decommissioning and disposal of wastes. As early power and test reactors have been decommissioned difficulties with end-of-the-fuel-cycle activities have become evident. Even the small test reactors common at the INEEL were not designed to facilitate their eventual decontamination, decommissioning, and dismantlement. The results are that decommissioning of these facilities is expensive, time consuming, relatively hazardous, and generates large volumes of waste. This situation clearly supports critics concerns about building a new generation of power reactors.

Richard H. Meservey

2006-06-01T23:59:59.000Z

413

Record of Decision; Continued operation of K, L, and P Reactors, Savannah River Site, Aiken, South Carolina  

SciTech Connect

The US Department of Energy (DOE) has considered the environmental impacts, benefits and costs, and institutional and programmatic needs associated with continued operation of the Savannah River Site (SRS) reactors, and has decided that it will continue to operate K and L Reactors at SRS, and will terminate operation of P Reactor in the immediate future and maintain it in cold standby. For P Reactor, this will involve the reactor's defueling; storage of its heavy water moderator in tanks in the reactor building; shutdown of reactor equipment and systems in a protected condition to prevent deterioration; and maintenance of the reactor in a defueled, protected status by a skeleton staff, which would permit any future decision to refuel and restart. Currently committed and planned upgrade activities will be discontinued for P Reactor. DOE will proceed with the safety upgrades and management system improvements currently scheduled for K Reactor in its program to satisfy the criteria of the Safety Evaluation Report (SER), and will conduct an Operational Readiness Review (ORR). The satisfaction of the SER criteria and completion of the ORR will demonstrate that the safety and health criteria for the resumption of production have been met. Reactor restart is expected to be in the third quarter of 1991 for K Reactor.

1991-01-01T23:59:59.000Z

414

Management Overview  

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

Small Modular Reactors Small Modular Reactors Materials Activities William Corwin Department of Energy Office of Nuclear Energy Office of Advanced Reactor Technologies DOE-NE Materials Coordination Webinar July 30, 2013 SMR Materials Work Packages Address Several Small Advanced Reactor Systems WP No. SMR R&D Program WBS Element Description Remarks SR-13IN060106 High Temperature Design Methodology - INL VHTR, Alloy 617 Code Case (with additional funding from NGNP) SR-13OR060107 High Temperature Design Methodology - ORNL VHTR, Alloy 617 Code Case SR-13AN060101 Materials for Advanced SMR Concepts - ANL Design methodology, SFR materials SR-13OR060103 Materials for Advanced SMR Concepts - ORNL Design methodology, SFR materials SR-13LA060102 Materials and Component Studies for Lead-

415

Acceptability of reactors in space  

SciTech Connect

Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it does not appear that reactors add measurably to the risk associated with the Space Transportation System.

Buden, D.

1981-01-01T23:59:59.000Z

416

Acceptability of reactors in space  

SciTech Connect

Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.

Buden, D.

1981-04-01T23:59:59.000Z

417

Thermonuclear Reflect AB-Reactor  

E-Print Network (OSTI)

The author offers a new kind of thermonuclear reflect reactor. The remarkable feature of this new reactor is a three net AB reflector, which confines the high temperature plasma. The plasma loses part of its energy when it contacts with the net but this loss can be compensated by an additional permanent plasma heating. When the plasma is rarefied (has a small density), the heat flow to the AB reflector is not large and the temperature in the triple reflector net is lower than 2000 - 3000 K. This offered AB-reactor has significantly less power then the currently contemplated power reactors with magnetic or inertial confinement (hundreds-thousands of kW, not millions of kW). But it is enough for many vehicles and ships and particularly valuable for tunnelers, subs and space apparatus, where air to burn chemical fuel is at a premium or simply not available. The author has made a number of innovations in this reactor, researched its theory, developed methods of computation, made a sample computation of typical project. The main point of preference for the offered reactor is its likely cheapness as a power source. Key words: Micro-thermonuclear reactor, Multi-reflex AB-thermonuclear reactor, Self-magnetic AB-thermonuclear reactor, aerospace thermonuclear engine.

Alexander Bolonkin

2008-03-26T23:59:59.000Z

418

NEUTRONIC REACTOR CHARGING AND DISCHARGING  

DOE Patents (OSTI)

A method and arrangement is presented for removing a fuel element from a neutronic reactor tube through which a liquid coolant is being circulaled. The fuel element is moved into a section of the tube beyond the reactor proper, and then the coolant in the tube between the fuel element and the reactor proper is frozen, so that the fuel element may be removed from the tube without loss of the coolant therein. The method is particularly useful in the case of a liquid metal- cooled reactor.

Zinn, W.H.

1959-07-14T23:59:59.000Z

419

Reactor & Nuclear Systems Publications | ORNL  

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

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

420

Mirror Advanced Reactor Study (MARS)  

DOE Green Energy (OSTI)

Progress in a two year study of a 1200 MWe commercial tandem mirror reactor (MARS - Mirror Advanced Reactor Study) has reached the point where major reactor system technologies are identified. New design features of the magnets, blankets, plug heating systems and direct converter are described. With the innovation of radial drift pumping to maintain low plug density, reactor recirculating power fraction is reduced to 20%. Dominance of radial ion and impurity losses into the halo permits gridless, circular direct converters to be dramatically reduced in size. Comparisons of MARS with the Starfire tokamak design are made.

Logan, B.G.

1983-03-28T23:59:59.000Z

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


421

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

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

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC)

422

Current Reactor Physics Benchmark Activities at the Idaho National Laboratory  

SciTech Connect

The International Reactor Physics Experiment Evaluation Project (IRPhEP) [1] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) [2] were established to preserve integral reactor physics and criticality experiment data for present and future research. These valuable assets provide the basis for recording, developing, and validating our integral nuclear data, and experimental and computational methods. These projects are managed through the Idaho National Laboratory (INL) and the Organisation for Economic Co-operation and Development Nuclear Energy Agency (OECD-NEA). Staff and students at the Department of Energy - Idaho (DOE-ID) and INL are engaged in the development of benchmarks to support ongoing research activities. These benchmarks include reactors or assemblies that support Next Generation Nuclear Plant (NGNP) research, space nuclear Fission Surface Power System (FSPS) design validation, and currently operational facilities in Southeastern Idaho.

John D. Bess; Margaret A. Marshall; Mackenzie L. Gorham; Joseph Christensen; James C. Turnbull; Kim Clark

2011-11-01T23:59:59.000Z

423

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

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

(RISMC) Advanced Test (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for

424

Transport Reactor Facility  

SciTech Connect

The Morgantown Energy Technology Center (METC) is currently evaluating hot gas desulfurization (HGD)in its on-site transport reactor facility (TRF). This facility was originally constructed in the early 1980s to explore advanced gasification processes with an entrained reactor, and has recently been modified to incorporate a transport riser reactor. The TRF supports Integrated Gasification Combined Cycle (IGCC) power systems, one of METC`s advanced power generation systems. The HGD subsystem is a key developmental item in reducing the cost and increasing the efficiency of the IGCC concept. The TRF is a unique facility with high-temperature, high-pressure, and multiple reactant gas composition capability. The TRF can be configured for reacting a single flow pass of gas and solids using a variety of gases. The gas input system allows six different gas inputs to be mixed and heated before entering the reaction zones. Current configurations allow the use of air, carbon dioxide, carbon monoxide, hydrogen, hydrogen sulfide, methane, nitrogen, oxygen, steam, or any mixture of these gases. Construction plans include the addition of a coal gas input line. This line will bring hot coal gas from the existing Fluidized-Bed Gasifier (FBG) via the Modular Gas Cleanup Rig (MGCR) after filtering out particulates with ceramic candle filters. Solids can be fed either by a rotary pocket feeder or a screw feeder. Particle sizes may range from 70 to 150 micrometers. Both feeders have a hopper that can hold enough solid for fairly lengthy tests at the higher feed rates, thus eliminating the need for lockhopper transfers during operation.

Berry, D.A.; Shoemaker, S.A.

1996-12-31T23:59:59.000Z

425

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

426

When Do Commercial Reactors Permanently Shut Down?  

Reports and Publications (EIA)

For those wishing to obtain current data, the following resources are available: U.S. reactors, go to EIA's nuclear reactor shutdown list. (Note: As of April 30, 2010, the last U.S. reactor to permanently shut down was Big Rock Point in 1997.) Foreign Reactors, go to the Power Reactor Information System (PRIS) on the International Atomic Energy Agency's website.

Information Center

2011-05-10T23:59:59.000Z

427

Reactor refueling containment system  

DOE Patents (OSTI)

A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

Gillett, J.E.; Meuschke, R.E.

1995-05-02T23:59:59.000Z

428

FOOD IRRADIATION REACTOR  

DOE Patents (OSTI)

An irradiation apparatus is described. It comprises a pressure vessel, a neutronic reactor active portion having a substantially greater height than diameter in the pressure vessel, an annular tank surrounding and spaced from the pressure vessel containing an aqueous indium/sup 1//sup 1//sup 5/ sulfate solution of approximately 600 grams per liter concentration, means for circulating separate coolants through the active portion and the space between the annular tank and the pressure vessel, radiator means adapted to receive the materials to be irradiated, and means for flowing the indium/sup 1//sup 1//sup 5/ sulfate solution through the radiator means.

Leyse, C.F.; Putnam, G.E.

1961-05-01T23:59:59.000Z

429

TRIGA reactor operating experience  

SciTech Connect

The Oregon State TRIGA Reactor (OSTR) has been in operation 3 years. Last August it was upgraded from 250 kW to 1000 kW. This was accomplished with little difficulty. During the 3 years of operation no major problems have been experienced. Most of the problems have been minor in nature and easily corrected. They came from lazy susan (dry bearing), Westronics Recorder (dead spots in the range), The Reg Rod Magnet Lead-in Circuit (a new type lead-in wire that does not require the lead-in cord to coil during rod withdrawal hss been delivered, much better than the original) and other small corrections.

Anderson, T.V. [Oregon State University, Corvallis, OR (United States)

1970-07-01T23:59:59.000Z

430

NEUTRONIC REACTOR CONSTRUCTION  

DOE Patents (OSTI)

A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

Vernon, H.C.; Goett, J.J.

1958-09-01T23:59:59.000Z

431

High flux reactor  

DOE Patents (OSTI)

A high flux reactor is comprised of a core which is divided into two symetric segments housed in a pressure vessel. The core segments include at least one radial fuel plate. The spacing between the plates functions as a coolant flow channel. The core segments are spaced axially apart such that a coolant mixing plenum is formed between them. A channel is provided such that a portion of the coolant bypasses the first core section and goes directly into the mixing plenum. The outlet coolant from the first core segment is mixed with the bypass coolant resulting in a lower inlet temperature to the lower core segment.

Lake, James A. (Idaho Falls, ID); Heath, Russell L. (Idaho Falls, ID); Liebenthal, John L. (Idaho Falls, ID); DeBoisblanc, Deslonde R. (Summit, NJ); Leyse, Carl F. (Idaho Falls, ID); Parsons, Kent (Idaho Falls, ID); Ryskamp, John M. (Idaho Falls, ID); Wadkins, Robert P. (Idaho Falls, ID); Harker, Yale D. (Idaho Falls, ID); Fillmore, Gary N. (Idaho Falls, ID); Oh, Chang H. (Idaho Falls, ID)

1988-01-01T23:59:59.000Z

432

Nuclear reactor control apparatus  

DOE Patents (OSTI)

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

Sridhar, Bettadapur N. (Cupertino, CA)

1983-11-01T23:59:59.000Z

433

Reactor refueling containment system  

DOE Patents (OSTI)

This report describes a method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

Gillett, J.E.; Meuschke, R.E.

1992-12-31T23:59:59.000Z

434

Biparticle fluidized bed reactor  

DOE Patents (OSTI)

A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase is described. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figures.

Scott, C.D.

1993-12-14T23:59:59.000Z

435

FAST NEUTRONIC REACTOR  

DOE Patents (OSTI)

This patent relates to a reactor and process for carrying out a controlled fast neutron chain reaction. A cubical reactive mass, weighing at least 920 metric tons, of uranium metal containing predominantly U/sup 238/ and having a U/sup 235/ content of at least 7.63% is assembled and the maximum neutron reproduction ratio is limited to not substantially over 1.01 by insertion and removal of a varying amount of boron, the reactive mass being substantially freed of moderator.

Snell, A.H.

1957-12-01T23:59:59.000Z

436

FUEL ASSAY REACTOR  

DOE Patents (OSTI)

A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

1962-12-25T23:59:59.000Z

437

Biparticle fluidized bed reactor  

DOE Patents (OSTI)

A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves.

Scott, Charles D. (Oak Ridge, TN)

1993-01-01T23:59:59.000Z

438

Biparticle fluidized bed reactor  

DOE Patents (OSTI)

A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves.

Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

1996-01-01T23:59:59.000Z

439

Biparticle fluidized bed reactor  

DOE Patents (OSTI)

A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves.

Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

1995-01-01T23:59:59.000Z

440

The Westinghouse Electric Corporation Reactor Vessel Radiation Surveillance Program  

Science Conference Proceedings (OSTI)

Westinghouse recognized that the disruption of the atomic lattice of metals by collision from energetic neutrons could alter the properties of the metals to such an extent that the changes could be of engineering significance. Furthermore, it was recognized that a physical-metallurgical phenomenon such as aging, both thermal and mechanical, also could alter the properties of a metal over its service life. Because of the potential changes in properties, reactor vessel radiation surveillance programs to monitor the effect of neutron radiation and other environmental factors on the reactor vessel materials during operational conditions over the life of the plant were initiated for Westinghouse plants with the insertion of reactor vessel material radiation surveillance capsules into the Yankee Atomic Company's Yankee Rowe plant in 1961.

Mayer, T.R.; Anderson, S.L.; Yanichko, S.E.

1983-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor aging management" 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.


441

Reactor closure design for a pool-type fast reactor  

SciTech Connect

The reactor closure is the topmost structural part of a reactor module. For a pool-type fast reactor it is an especially important structure because it provides the interface between the primary coolant system and the main access area above the closure. The reactor closure comprises a stationary deck, a rotatable plug, the boundary elements of primary system and containment penetrations for equipment and auxiliary systems. This paper evaluates two different reactor closure design concepts, referred to as ''warm'' deck and ''hot'' deck, for a pool-type fast reactor with respect to their design features, technical merits, and economic benefits. The evaluation also includes functional, structural, and thermal analyses of the two deck design concepts. Issues related to their fabrication and shipping to the plant site are also addressed. The warm deck is a thick solid steel plate with under-the-deck insulation consisting of many layers of steel plates. The hot deck is a box-type structure consisting of a bottom plate reinforced with vertical ribs and cylinders. For insulation and radiation shielding, the region of the hot deck above the bottom plate is filled with steel balls. Conventional insulation is added on the top to further reduce heat loss into area above the deck. The design choice of the closure deck is strongly dependent on design features of the reactor; especially on the reactor module support. While the warm deck is preferable with the top support, the hot deck is better suited for the bottom support design of the module.

Chung, H.; Seidensticker, R.W.; Kann, W.J.; Bump, T.R.; Schatmeier, C.

1986-01-01T23:59:59.000Z

442

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect

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

Not Available

1991-04-01T23:59:59.000Z

443

Fast Reactor Fuel Type and Reactor Safety Performance  

Science Conference Proceedings (OSTI)

Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and thermophysical properties of the fuel and their compatibility with the reactor coolant, with corresponding differences in the challenges presented to the reactor developers. Accident phenomena are discussed for the sodium-cooled fast reactor based on the mechanistic progression of conditions from accident initiation to accident termination, whether a benign state is achieved or more severe consequences are expected. General principles connecting accident phenomena and fuel properties are developed from the oxide and metal fuel safety analyses, providing guidelines that can be used as part of the evaluation for selection of fuel type for the sodium-cooled fast reactor.

R. Wigeland; J. Cahalan

2009-09-01T23:59:59.000Z

444

Aging and Holography  

E-Print Network (OSTI)

Aging phenomena are examples of `non-equilibrium criticality' and can be exemplified by systems with Galilean and scaling symmetries but no time translation invariance. We realize aging holographically using a deformation of a non-relativistic version of gauge/gravity duality. Correlation functions of scalar operators are computed using holographic real-time techniques, and agree with field theory expectations. At least in this setup, general aging phenomena are reproduced holographically by complexifying the bulk space-time geometry, even in Lorentzian signature.

Juan I. Jottar; Robert G. Leigh; Djordje Minic; Leopoldo A. Pando Zayas

2010-04-21T23:59:59.000Z

445

Finite Temperature Aging Holography  

E-Print Network (OSTI)

We construct the gravity background which describes the dual field theory with aging invariance. We choose the decay modes of the bulk scalar field in the internal spectator direction to obtain the dissipative behavior of the boundary correlation functions of the dual scalar fields. In particular, the two-time correlation function at zero temperature has the characteristic features of the aging system: power law decay, broken time translation and dynamical scaling. We also construct the black hole backgrounds with asymptotic aging invariance. We extensively study characteristic behavior of the finite temperature two-point correlation function via analytic and numerical methods.

Seungjoon Hyun; Jaehoon Jeong; Bom Soo Kim

2011-08-29T23:59:59.000Z

446

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Assessment of High Value Surveillance Materials Assessment of High Value Surveillance Materials Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely

447

Advanced Burner Reactor Preliminary NEPA Data Study.  

Science Conference Proceedings (OSTI)

The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly documents the extensive evaluation which was performed on the anticipated environmental impacts of that plant. This source can be referenced in the open literature and is publicly available. The CRBRP design was also of a commercial demonstration plant size - 975 MWth - which falls in the middle of the range of ABR plant sizes being considered (250 MWth to 2000 MWth). At the time the project was cancelled, the CRBRP had progressed to the point of having completed the licensing application to the Nuclear Regulatory Commission (NRC) and was in the process of receiving NRC approval. Therefore, it was felt that [CRBRP, 1977] provides some of the best available data and information as input to the GNEP PEIS work. CRBRP was not the source of all the information in this document. It is also expected that the CRBRP data will be bounding from the standpoint of commodity usage because fast reactor vendors will develop designs which will focus on commodity and footprint reduction to reduce the overall cost per kilowatt electric compared with the CRBR plant. Other sources used for this datacall information package are explained throughout this document and in Appendix A. In particular, see Table A.1 for a summary of the data sources used to generate the datacall information.

Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R.; Kim, T. K.; Yang, W. S.; Nuclear Engineering Division

2007-10-15T23:59:59.000Z

448

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

National Nuclear Security Administration (NNSA)

Domestic U.S. Reactor Conversions: Fact Sheet | National Nuclear Security Domestic U.S. Reactor Conversions: Fact Sheet | 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 > Fact Sheets > Domestic U.S. Reactor Conversions: Fact Sheet Fact Sheet Domestic U.S. Reactor Conversions: Fact Sheet Mar 23, 2012 The National Nuclear Security Administration (NNSA) helps convert research

449

Cable Reliability Management Strategies  

Science Conference Proceedings (OSTI)

In previous collaborative research with several utilities, EPRI developed methods for forecasting the failures and cash flows associated with managing a population of underground cable. This study builds on that research and provides a method for developing least-cost repair and replacement policies for such a population. The policies specify when to replace a cable section or an entire circuit, depending on the age of the section and the number of failures the section has experienced.

2003-12-10T23:59:59.000Z

450

Reactor Operations informal monthly report December 1994  

Science Conference Proceedings (OSTI)

Reactor operations at the MRR and HFBR reactors at Brookhaven National Laboratory are presented for December 1994. Reactor run-time and power levels, instrumentation, mechanical maintenance, occurrence reports, and safety information are included.

NONE

1994-12-01T23:59:59.000Z

451

Measurement of theta13 with reactor neutrinos  

E-Print Network (OSTI)

? detectors ~1 km nuclear reactor Figure 2. Concept of a 2-Measurement of ? 13 with Reactor Neutrinos K.M. Heeger a ,power plant, a future reactor neutrino experiment has the

Heeger, Karsten M.; Freedman, Stuart J.; Kadel, Richard W.; Luk, Kam-Biu

2004-01-01T23:59:59.000Z

452

Granular Dynamics in Pebble Bed Reactor Cores  

E-Print Network (OSTI)

in a pebble-bed nuclear reactor,” Phys. Rev. E, vol. 74, no.cycles of the pebble bed reactor,” Nuclear Engineering andoptimization of pebble-bed reactors,” Annals of Nuclear

Laufer, Michael Robert

2013-01-01T23:59:59.000Z

453

Granular Dynamics in Pebble Bed Reactor Cores  

E-Print Network (OSTI)

flow in a pebble-bed nuclear reactor,” Phys. Rev. E, vol.from the current fleet of nuclear reactors far outweigh thethrough the core of a nuclear reactor. This regime includes

Laufer, Michael Robert

2013-01-01T23:59:59.000Z

454

Measurement of theta13 with reactor neutrinos  

E-Print Network (OSTI)

? detectors ~1 km nuclear reactor Figure 2. Concept of a 2-in the core of a nuclear reactor. Re- actor ? e are usuallybetween 0.1- 3 km from a nuclear reactor to measure the rate

Heeger, Karsten M.; Freedman, Stuart J.; Kadel, Richard W.; Luk, Kam-Biu

2004-01-01T23:59:59.000Z

455

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...

456

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

457

HEU age determination  

SciTech Connect

A technique has been developed to determine the Highly Enriched Uranium (HEU) Age which is defined as the time since the HEU was produced in an enrichment process. The HEU age is determined from the ratios of relevant uranium parents and their daughters viz {sup 230}Th/{sup 234}U and {sup 231}Pa/{sup 235}U. Uranium isotopes are quantitatively measured by their characteristic gammas and their daughters by alpha spectroscopy. In some of the samples where HEU is enriched more than 99%, the only mode of HEU age determination is by the measurement of {sup 231}Pa since there is negligible quantity of {sup 230}Th due to very low atom concentrations of {sup 234}U in the sample. In this paper we have presented data and methodology of finding the age of two HEU samples.

Moorthy, A.R.; Kato, W.Y.

1995-08-01T23:59:59.000Z

458

Speech understanding and aging  

Science Conference Proceedings (OSTI)

The increasing age of the population forecasts a corresponding increase in the number of people suffering from hearing loss and who therefore have difficulty with speech communication. This article

Working Group on Speech Understanding and Aging

1988-01-01T23:59:59.000Z

459

Management Responsibilities  

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

Management Responsibilities Management Responsibilities Depleted UF6 Management Responsibilities DOE has responsibility for safe and efficient management of approximately 700,000 metric tons of depleted UF6. Organizational Responsibilities In the United States, the U.S. Department of Energy is responsible for managing all the depleted uranium that has been generated by the government and has been declared surplus to national defense needs. In addition, as a result of two memoranda of agreement that have been signed between the DOE and USEC, the DOE has assumed management responsibility for approximately 145,000 metric tons of depleted UF6 that has been or will be generated by USEC. Any additional depleted UF6 that USEC generates will be USEC's responsibility to manage. DOE Management Responsibility

460

Conceptual innovations in hybrid reactors  

SciTech Connect

A number of innovations in the conception of fusion-fission hybrid reactors, including the blanket, the fusion driver, the coupling of the fusion and the fission components as well as the application of hybrid reactors are described, and their feasibility assessed.

Greenspan, E.; Miley, G.H.

1980-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor aging management" 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.