Preparation for Future Defuelling and Decommissioning Works on EDF Energy's UK Fleet of Advanced Gas Cooled Reactors

International Nuclear Information System (INIS)

Bryers, John; Ashmead, Simon

2016-01-01

EDF Energy/Nuclear Generation is the owner and operator of 14 Advanced Gas cooled Reactors (AGR) and one Pressurised Water Reactor (PWR), on 8 nuclear stations in the UK. EDF Energy/Nuclear Generation is responsible for all the activities associated with the end of life of its nuclear installations: de-fuelling, decommissioning and waste management. As the first AGR is forecast to cease generation within 10 years, EDF Energy has started planning for the decommissioning. This paper covers: - broad outline of the technical strategy and arrangements for future de-fuelling and decommissioning works on the UK AGR fleet, - high level strategic drivers and alignment with wider UK nuclear policy, - overall programme of preparation and initial works, - technical approaches to be adopted during decommissioning. (authors)

Strategic vehicle fleet management - the replacement problem

Directory of Open Access Journals (Sweden)

Adam Redmer

2016-03-01

Full Text Available Background: Fleets constitute the most important production means in transportation. Their appropriate management is crucial for all companies having transportation duties. The paper is the third one of a series of three papers that the author dedicates to the strategic vehicle fleet management topic. Material and methods: The paper discusses ways of building replacement strategies for companies' fleets of vehicles. It means deciding for how long to exploit particular vehicles in a fleet (the fleet replacement problem - FR. The essence of this problem lies in the minimization of vehicle / fleet exploitation costs by balancing ownership and utilization costs and taking into account budget limitations. In the paper an original mathematical model (an optimization method allowing for the FR analysis is proposed. Results: An application of the proposed optimization method in a real-life decision situation (the case study within the Polish environment and the obtained solution are presented. The solution shows that there exist optimal exploitation periods of particular vehicles in a fleet. However, combination of them gives a replacement plan for an entire fleet violating budget constraints. But it is possible to adjust individual age to replacement of particular vehicles to fulfill budget constraints without losing economical optimality of a developed replacement plan for an entire fleet. Conclusions: The paper is the last one of a series of three papers that the author dedicated to the strategic vehicle fleet management topic including the following managerial decision problems: MAKE-or-BUY, sizing / composition and replacement.

Strain components of nuclear-reactor-type concretes during first heat cycle

International Nuclear Information System (INIS)

Khoury, G.A.

1995-01-01

Strains of three advanced-gas-cooled-reactor-type nuclear reactor concretes were measured during the first heat cycle and their relative thermal stability determined. It was possible to isolate for the first time the shrinkage component for the period during heating. Predictions of the residual strains for the loaded specimens can be made by simple superposition of creep and shrinkage components up to a certain critical temperature, which for basalt concrete is about 500 C and for limestone concrete is about 200-300 C. Above the critical temperature, an expansive ''cracking'' strain component is present. It is shown that the strain behaviour of concrete provides a sensitive indication of its thermal stability during heating and subsequent cooling. (orig.)

Manufacture of components for Canadian reactor programs

International Nuclear Information System (INIS)

Perry, L.P.

Design features, especially those relating to calandrias, are pointed out for many CANDU-type reactors and the Taiwan research reactor. The special requirements shouldered by the Canadian suppliers of heavy reactor components are analyzed. (E.C.B.)

New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

International Nuclear Information System (INIS)

Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

1996-01-01

The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

Application of Strategic Planning Process with Fleet Level Analysis Methods

Science.gov (United States)

Mavris, Dimitri N.; Pfaender, Holger; Jimenez, Hernando; Garcia, Elena; Feron, Eric; Bernardo, Jose

2016-01-01

The goal of this work is to quantify and characterize the potential system-wide reduction of fuel consumption and corresponding CO2 emissions, resulting from the introduction of N+2 aircraft technologies and concepts into the fleet. Although NASA goals for this timeframe are referenced against a large twin aisle aircraft we consider their application across all vehicle classes of the commercial aircraft fleet, from regional jets to very large aircraft. In this work the authors describe and discuss the formulation and implementation of the fleet assessment by addressing the main analytical components: forecasting, operations allocation, fleet retirement, fleet replacement, and environmental performance modeling.

CleanFleet. Final report: Executive summary

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-12-01

CleanFleet, formally known as the South Coast Alternative Fuels Demonstration, was a comprehensive demonstration of alternative fuel vehicles (AFVs) in daily commercial service. Between April 1992 and September 1994, five alternative fuels were tested in 84 panel vans: compressed natural gas (CNG), propane gas, methanol as M-85, California Phase 2 reformulated gasoline (RFG), and electricity. The AFVs were used in normal FedEx package delivery service in the Los Angeles basin alongside 27 {open_quotes}control{close_quotes} vans operating on regular gasoline. The liquid and gaseous fuel vans were model year 1992 vans from Ford, Chevrolet, and Dodge. The two electric vehicles (EVs) were on loan to FedEx from Southern California Edison. The AFVs represented a snapshot in time of 1992 technologies that (1) could be used reliably in daily FedEx operations and (2) were supported by the original equipment manufacturers (OEMs). A typical van is shown in Figure 2. The objective of the project was to demonstrate and document the operational, emissions, and economic status of alternative fuel, commercial fleet delivery vans in the early 1990s for meeting air quality regulations in the mid to late 1990s. During the two-year demonstration, CleanFleet`s 111 vehicles travelled more than three million miles and provided comprehensive data on three major topics: fleet operations, emissions, and fleet economics. Fleet operations were examined in detail to uncover and resolve problems with the use of the fuels and vehicles in daily delivery service. Exhaust and evaporative emissions were measured on a subset of vans as they accumulated mileage. The California Air Resources Board (ARB) measured emissions to document the environmental benefits of these AFVs. At the same time, CleanFleet experience was used to estimate the costs to a fleet operator using AFVs to achieve the environmental benefits of reduced emissions.

Multiobjective optimization for nuclear fleet evolution scenarios using COSI

Directory of Open Access Journals (Sweden)

Freynet David

2016-01-01

Full Text Available The consequences of various fleet evolution options on material inventories and flux in fuel cycle and waste can be analysed by means of transition scenario studies. The COSI code is currently simulating chronologically scenarios whose parameters are fully defined by the user and is coupled with the CESAR depletion code. As the interactions among reactors and fuel cycle facilities can be complex, and the ways in which they may be configured are many, the development of optimization methodology could improve scenario studies. The optimization problem definition needs to list: (i criteria (e.g. saving natural resources and minimizing waste production; (ii variables (scenario parameters related to reprocessing, reactor operation, installed power distribution, etc.; (iii constraints making scenarios industrially feasible. The large number of scenario calculations needed to solve an optimization problem can be time-consuming and hardly achievable; therefore, it requires the shortening of the COSI computation time. Given that CESAR depletion calculations represent about 95% of this computation time, CESAR surrogate models have been developed and coupled with COSI. Different regression models are compared to estimate CESAR outputs: first- and second-order polynomial regressions, Gaussian process and artificial neural network. This paper is about a first optimization study of a transition scenario from the current French nuclear fleet to a Sodium Fast Reactors fleet as defined in the frame of the 2006 French Act for waste management. The present article deals with obtaining the optimal scenarios and validating the methodology implemented, i.e. the coupling between the simulation software COSI, depletion surrogate models and a genetic algorithm optimization method.

Fleet DNA Brings Fleet Data to Life, Informs R&D | NREL

Science.gov (United States)

Fleet DNA Brings Fleet Data to Life, Informs R&D Science and Technology Highlights Highlights in Research & Development Fleet DNA Brings Fleet Data to Life, Informs R&D Key Research Results Achievement Built on over 11.5 million miles of vehicle operations data, Fleet DNA helps users

Energy deposition in STARFIRE reactor components

International Nuclear Information System (INIS)

Gohar, Y.; Brooks, J.N.

1985-04-01

The energy deposition in the STARFIRE commercial tokamak reactor was calculated based on detailed models for the different reactor components. The heat deposition and the 14 MeV neutron flux poloidal distributions in the first wall were obtained. The poloidal surface heat load distribution in the first wall was calculated from the plasma radiation. The Monte Carlo method was used for the calculation to allow an accurate modeling for the reactor geometry

Fleet DNA Brings Fleet Data to Life, Informs R&D | News | NREL

Science.gov (United States)

Fleet DNA Brings Fleet Data to Life, Informs R&D Fleet DNA Brings Fleet Data to Life, Informs R De La Rosa, NREL 34672 The Fleet DNA clearinghouse of commercial vehicle operations data features Odyne-have tapped into Fleet DNA." The data-driven insight and decision-making capabilities

Strategic vehicle fleet management - the composition problem

Directory of Open Access Journals (Sweden)

Adam Redmer

2015-03-01

Full Text Available Background: Fleets constitute the most important production means in transportation. Their appropriate management is crucial for all companies having transportation duties. The paper is the second one of a series of three papers that the author dedicates to the strategic vehicle fleet management topic. Material and methods: The paper discusses ways of building companies' fleets of vehicles. It means deciding on the number of vehicles in a fleet (the fleet sizing problem - FS and types of vehicles in a fleet (the fleet composition problem - FC. The essence of both problems lies in balancing transportation supply and demand taking into account different demand types to be fulfilled and different vehicle types that can be put into a fleet. Vehicles, which can substitute each other while fulfilling different demand types. In the paper an original mathematical model (an optimization method allowing for the FS/FC analysis is proposed. Results: An application of the proposed optimization method in a real-life decision situation (the case study within the Polish environment and the obtained solution are presented. The solution shows that there exist some best fitted (optimal fleet size / composition matching company's transportation requirements. An optimal fleet size / composition allows for a significantly higher fleet utilization (10-15% higher than any other, including random fleet structure. Moreover, any changes in the optimal fleet size / composition, even small ones, result in a lower utilization of vehicles (lower by a few percent. Conclusions: The presented in this paper analysis, on the one hand, is consistent with a widespread opinion that the number of vehicle types in a fleet should be limited. In the other words it means that the versatility / interchangeability of vehicles is very important. On the other hand, the analysis proves that even small changes in a fleet size / fleet composition can result in an important changes of the fleet

Aging of metal components in US nuclear reactors

International Nuclear Information System (INIS)

Mayfield, M.E.; Strosnider, J.R.

1998-01-01

This paper presents an overview of the aging of metal components in U.S. Light Water Reactors. The types of degradation being experienced in components such as the pressure vessel, piping, reactor internals, and steam generators, and the programs being implemented to manage the degradation are discussed. (author)

FleetDASH

Energy Technology Data Exchange (ETDEWEB)

Singer, Mark R [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

2017-09-06

FleetDASH helps federal fleet managers maximize their use of alternative fuel. This presentation explains how the dashboard works and demonstrates the newest capabilities added to the tool. It also reviews complementary online tools available to fleet managers on the Alternative Fuel Data Center.

Characterization of the Three Mile Island Unit-2 reactor building atmosphere prior to the reactor building purge

International Nuclear Information System (INIS)

Hartwell, J.K.; Mandler, J.W.; Duce, S.W.; Motes, B.G.

1981-05-01

The Three Mile Island Unit-2 reactor building atmosphere was sampled prior to the reactor building purge. Samples of the containment atmosphere were obtained using specialized sampling equipment installed through penetration R-626 at the 358-foot (109-meter) level of the TMI-2 reactor building. The samples were subsequently analyzed for radionuclide concentration and for gaseous molecular components (O 2 , N 2 , etc.) by two independent laboratories at the Idaho National Engineering Laboratory (INEL). The sampling procedures, analysis methods, and results are summarized

Component failures that lead to reactor scrams

International Nuclear Information System (INIS)

Burns, E.T.; Wilson, R.J.; Lim, E.Y.

1980-04-01

This report summarizes the operating experience scram data compiled from 35 operating US light water reactors (LWRs) to identify the principal components/systems related to reactor scrams. The data base utilized to identify the scram causes is developed from a EPRI-utility sponsored survey conducted by SAI coupled with recent data from the USNRC Gray Books. The reactor population considered in this evaluation is limited to 23 PWRs and 12 BWRs because of the limited scope of the program. The population includes all the US NSSS vendors. It is judged that this population accurately characterizes the component-related scrams in LWRs over the first 10 years of plant operation

Ageing management program for reactor components in HANARO

International Nuclear Information System (INIS)

Cho, Yeong-Garp; Wu, Sang-Ik; Lee, Jung-Hee; Ryu, Jeong-Soo; Park, Yong-Chul; Wu, Jong-Sup; Jun, Byung Jin

2003-01-01

The HANARO, an open-tank-in-pool type research reactor of 30MWth power in Korea, has operated for 8 years since its initial criticality in February of 1995. The reactor power has been gradually increased to 24 MWth through the service period. Therefore the reactor age is very young from the viewpoint of the ageing effect on the reactor structure and components by neutron irradiation considering the expected reactor lifetime. But, we have a few programs to manage the ageing from the aspect of design lifetime of reactor components. This paper summarizes the overall progress and plan for the ageing management for the reactor components including lifetime extension and design improvement, remote measurements and in-service inspections. The shutoff units and control absorber units have aged more rapidly than other structures or components because the number of rod drop cycles was higher than expected at the design stage. The system commissioning tests, periodic performance tests, and weekly operation for the stable supply of medical radioisotopes overriding the normal cycle operation have contributed to the high frequency of rod drop. Therefore, we have instituted a program to extend the lifetime of the shutoff units and the control absorber units. This program includes an endurance test to verify the performance for the extended number of drops and the management of shutdown methods to minimize the drop cycles for both the shutoff units and the control absorber units. The program also includes the design improvement of the damper mechanism of the control absorber units to reduce the impact force caused by rod drop. The inner shell of the reflector vessel surrounding the core is the most critical part from the viewpoint of neutron irradiation. The periodic measurement of the dimensional change in the vertical straightness of the inner shell is considered as one of the in-service inspections. We developed a few tools and verified the performance to measure the

Multi-component controllers in reactor physics optimality analysis

International Nuclear Information System (INIS)

Aldemir, T.

1978-01-01

An algorithm is developed for the optimality analysis of thermal reactor assemblies with multi-component control vectors. The neutronics of the system under consideration is assumed to be described by the two-group diffusion equations and constraints are imposed upon the state and control variables. It is shown that if the problem is such that the differential and algebraic equations describing the system can be cast into a linear form via a change of variables, the optimal control components are piecewise constant functions and the global optimal controller can be determined by investigating the properties of the influence functions. Two specific problems are solved utilizing this approach. A thermal reactor consisting of fuel, burnable poison and moderator is found to yield maximal power when the assembly consists of two poison zones and the power density is constant throughout the assembly. It is shown that certain variational relations have to be considered to maintain the activeness of the system equations as differential constraints. The problem of determining the maximum initial breeding ratio for a thermal reactor is solved by treating the fertile and fissile material absorption densities as controllers. The optimal core configurations are found to consist of three fuel zones for a bare assembly and two fuel zones for a reflected assembly. The optimum fissile material density is determined to be inversely proportional to the thermal flux

UK fast reactor components - sodium removal decontamination and requalification

Energy Technology Data Exchange (ETDEWEB)

Donaldson, D M [FRDD, UKAEA, Risley (United Kingdom); Bray, J A; Newson, I H [UKAEA, Dounreay Nuclear Power Establishment, Thurso (United Kingdom)

1978-08-01

Over the past two decades extensive experience on sodium removal techniques has been gained at the UKAEA's Dounreay Nuclear Establishment from both the Dounreay Fact Reactor (DFR) and the Prototype Fast Reactor (PFR). This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Part 2 of the paper, which describes recent operations associated with the PFR, demonstrates the background to these views. This past and continuing experience is being used in forming the basis of the plant to be provided for sodium removal, decontamination and requalification of components in the UK's future commercial fast reactors. Further improvements in techniques and in component designs can be expected in the course of the next few years. Consequently UK philosophy and approach with respect to maintenance and repair operations is sufficiently flexible to enable relevant improvements to be incorporated into the next scheduled fast reactor - the Commercial Demonstration Fast Reactor (CUR). This paper summarises the factors which are being taken into consideration in this continuously advancing field.

UK fast reactor components - sodium removal decontamination and requalification

International Nuclear Information System (INIS)

Donaldson, D.M.; Bray, J.A.; Newson, I.H.

1978-01-01

Over the past two decades extensive experience on sodium removal techniques has been gained at the UKAEA's Dounreay Nuclear Establishment from both the Dounreay Fact Reactor (DFR) and the Prototype Fast Reactor (PFR). This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Part 2 of the paper, which describes recent operations associated with the PFR, demonstrates the background to these views. This past and continuing experience is being used in forming the basis of the plant to be provided for sodium removal, decontamination and requalification of components in the UK's future commercial fast reactors. Further improvements in techniques and in component designs can be expected in the course of the next few years. Consequently UK philosophy and approach with respect to maintenance and repair operations is sufficiently flexible to enable relevant improvements to be incorporated into the next scheduled fast reactor - the Commercial Demonstration Fast Reactor (CUR). This paper summarises the factors which are being taken into consideration in this continuously advancing field

Simulating the behaviour of zirconium-alloy components in nuclear reactors

International Nuclear Information System (INIS)

Coleman, C.E.

2001-12-01

To prevent failure in nuclear components one needs to understand the interactions between adjacent materials and the changes in their physical properties during all phases of reactor operation. Three examples from CANDU reactors are described to illustrate the use of simulations that imitate complicated reactor situations. These are: swelling tests that led to a method for increasing the tolerance or Zircaloy fuel cladding to power ramps; observations of the behaviour of leaking cracks in Zr-2.5Nb pressure tubes that provide confidence in the use of leak-before-break as part of the defence against flaw development; and contact boiling tests on modifications to the surfaces of Zircaloy calandria tubes that enhance the ability of the heavy water moderator to act as a heat sink after a postulated loss-of-coolant accident. (author)

Gap and impact of LMR [Liquid Metal Reactor] piping systems and reactor components

International Nuclear Information System (INIS)

Ma, D.C.; Gvildys, J.; Chang, Y.W.

1987-01-01

Because of high operation temperature, the LMR (Liquid Metal Reactor) plant is characterized by the thin-walled piping and components. Gaps are often present to allow free thermal expansion during normal plant operation. Under dynamic loadings, such as seismic excitation, if the relative displacement between the components exceeds the gap distance, impacts will occur. Since the components and piping become brittle over their design lifetime, impact is of important concern for it may lead to fractures of components and other serious effects. This paper deals with gap and impact problems in the LMR reactor components and piping systems. Emphasis is on the impacts due to seismic motion. Eight sections are contained in this paper. The gap and impact problems in LMR piping systems are described and a parametric study is performed on the effects of gap-induced support nonlinearity on the dynamics characteristics of the LMR piping systems. Gap and impact problems in the LMR reactor components are identified and their mathematical models are illustrated, and the gap and impact problems in the seismic reactor scram are discussed. The mathematical treatments of various impact models are also described. The uncertainties in the current seismic impact analyses of LMR components and structures are presented. An impact test on a 1/10-scale LMR thermal liner is described. The test results indicated that several clusters of natural modes can be excited by the impact force. The frequency content of the excited modes depends on the duration of the impact force; the shorter the duration, the higher the frequency content

Aging management of major light water reactor components

International Nuclear Information System (INIS)

Shah, V.N.; Sinha, U.P.; Ware, A.G.

1992-01-01

Review of technical literature and field experience has identified stress corrosion cracking as one of the major degradation mechanisms for the major light water reactor components. Three of the stress corrosion cracking mechanisms of current concern are (a) primary water stress corrosion cracking (PWSCC) in pressurized water reactors, and (b) intergranular stress corrosion cracking (IGSCC) and (c) irradiation-assisted stress corrosion cracking (IASCC) in boiling water reactors. Effective aging management of stress corrosion cracking mechanisms includes evaluation of interactions between design, materials, stressors, and environment; identification and ranking of susceptible sites; reliable inspection of any damage; assessment of damage rate; mitigation of damage; and repair and replacement using corrosion-resistant materials. Management of PWSCC includes use of lower operating temperatures, reduction in residual tensile stresses, development of reliable inspection techniques, and use of Alloy 690 as replacement material. Management of IGSCC of nozzle and attachment welds includes use of Alloy 82 as weld material, and potential use of hydrogen water chemistry. Management of IASCC also includes potential use of hydrogen water chemistry

Generic component reliability data for research reactor PSA

International Nuclear Information System (INIS)

1997-02-01

The purpose of this document is to provide reference generic component-reliability information for a variety of research reactor types. As noted in Section 2 and Table IV, component data accumulated over many years is in the database. It is expected that the report should provide representative data which will remain valid for a number of years. The database provides component failure rates on a time and/or demand related basis according to the operational modes of the components. No update of the database is presently planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available by the contributing research reactor facilities themselves. As noted in Section 1.1, the report does not include a detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussions regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, 7 tabs

Component failures at pressurized water reactors. Final report

International Nuclear Information System (INIS)

Reisinger, M.F.

1980-10-01

Objectives of this study were to identify those systems having major impact on safety and availability (i.e. to identify those systems and components whose failures have historically caused the greatest number of challenges to the reactor protective systems and which have resulted in greatest loss of electric generation time). These problems were identified for engineering solutions and recommendations made for areas and programs where research and development should be concentrated. The program was conducted in three major phases: Data Analysis, Engineering Evaluation, Cost Benefit Analysis

Draft of standard for graphite core components in high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

2010-01-01

For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

Ageing investigation and upgrading of components/systems of Kartini research reactor

Energy Technology Data Exchange (ETDEWEB)

Syarip,; Setiawan, Widi [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia)

1998-10-01

Kartini research reactor has been operated in good condition and has demonstrated successful operation for the past 18 years, utilized for: reactor kinetic and control studies, instrumentation tests, neutronic and thermohydraulic studies, routine neutron activation analysis, reactor safety studies, training for research reactor operators and supervisors, and reactor physics experiments. Several components of Kartini reactor use components from the abandoned IRT-2000 Project at Serpong and from Bandung Reactor Centre such as: reactor tank, reactor core, heat exchanger, motor blower for ventilation system, fuel elements, etc. To maintain a good operating performance and also for aging investigation purposes, the component failure data collection has been done. The method used is based on the Manual on Reliability Data Collection For Research Reactor PSAs, IAEA TECDOC 636, and analyzed by using Data Entry System (DES) computer code. Analysis result shows that the components/systems failure rate of Kartini reactor is around 1,5.10{sup -4} up to 2,8.10{sup -4} per hour, these values are within the ranges of the values indicated in IAEA TECDOC 478. Whereas from the analysis of irradiation history shows that the neutron fluence of fuel element with highest burn-up (2,05 gram U-235 in average) is around 1.04.10{sup 16} n Cm{sup -2} and this value is still far below its limiting value. Some reactor components/systems have been replaced and upgraded such as heat exchanger, instrumentation and control system (ICS), etc. The new reactor ICS was installed in 1994 which is designed as a distributed structure by using microprocessor based systems and bus system technology. The characteristic and operating performance of the new reactor ICS, as well as the operation history and improvement of the Kartini research reactor is presented. (J.P.N.)

Mechanical development for reliable reactor components

International Nuclear Information System (INIS)

Ross-Ross, P.A.; Metcalfe, R.

1983-09-01

The CANDU reactor has achieved worldwide distinction because of its reliable performance. To achieve this, special attention was given to the reliability and maintainability of components in the heavy water circuits. Development programs were initiated early in the history of the CANDU reactor to improve the effectiveness of pump seals, valves, and static seals because of unacceptable performance of the commercial equipment then available. As a result, pump seals with a five year life now appear achievable, and valves and static seals are no longer a significant concern in CANDU reactors. Increasing effort is being given remotely operated tools and fabrication systems for radioactive environments

Development of components for the gas-cooled fast breeder reactor program

International Nuclear Information System (INIS)

Dee, J.B.; Macken, T.

1977-01-01

The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core. The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs. (Auth.)

The Potential Impact of Collaborative and Three-Dimensional Imaging Technology on SHIPMAIN Fleet Modernization Plan

National Research Council Canada - National Science Library

Seaman, Nathan L; Housel, Thomas; Mun, Jonathan

2008-01-01

Maintenance and modernization of the US Navy fleet is big business. The Navy has invested substantial fiscal and human resources to standardize the processes used to accomplish maintenance modernization and repair for its fleet of ships...

Three dimensional diffusion calculations of nuclear reactors

International Nuclear Information System (INIS)

Caspo, N.

1981-07-01

This work deals with the three dimensional calculation of nuclear reactors using the code TRITON. The purposes of the work were to perform three-dimensional computations of the core of the Soreq nuclear reactor and of the power reactor ZION and to validate the TRITON code. Possible applications of the TRITON code in Soreq reactor calculations and in power reactor research are suggested. (H.K.)

Integrity of pressurized water electronuclear reactor vessels. The case of French reactors

International Nuclear Information System (INIS)

2012-01-01

This document aims at identifying elements related to design, manufacturing and control during operation of reactor vessels of the French electronuclear fleet, and more precisely as far as vessel ferrule is concerned. It briefly describes the typical design and elements of most of French PWR vessels with respect to the reactor type (900 MWe, 1300 MWe, 1450 MWe, EPR). It recalls some measures regarding design (for embrittlement assessment) and manufacturing processes (forging operations for shell fabrication, coatings). It discusses the different manufacturing defects which have been noticed (under the coatings, due to hydrogen, and intergranular loss of cohesion due to re-heating). It more particularly comments defects noticed on a Belgium power station reactor in Doel, defects due to hydrogen and some other defects noticed in the French reactor fleet. It presents the different types of control which are performed on vessel shells during operation

Numerical analysis of magnetoelastic coupled buckling of fusion reactor components

International Nuclear Information System (INIS)

Demachi, K.; Yoshida, Y.; Miya, K.

1994-01-01

For a tokamak fusion reactor, it is one of the most important subjects to establish the structural design in which its components can stand for strong magnetic force induced by plasma disruption. A number of magnetostructural analysis of the fusion reactor components were done recently. However, in these researches the structural behavior was calculated based on the small deformation theory where the nonlinearity was neglected. But it is known that some kinds of structures easily exceed the geometrical nonlinearity. In this paper, the deflection and the magnetoelastic buckling load of fusion reactor components during plasma disruption were calculated

Challenges in design of zirconium alloy reactor components

International Nuclear Information System (INIS)

Kakodkar, Anil; Sinha, R.K.

1992-01-01

Zirconium alloy components used in core-internal assemblies of heavy water reactors have to be designed under constraints imposed by need to have minimum mass, limitations of fabrication, welding and joining techniques with this material, and unique mechanisms for degradation of the operating performance of these components. These constraints manifest as challenges for design and development when the size, shape and dimensions of the components and assemblies are unconventional or untried, or when one is aiming for maximization of service life of these components under severe operating conditions. A number of such challenges were successfully met during the development of core-internal components and assemblies of Dhruva reactor. Some of the then untried ideas which were developed and successfully implemented include use of electron beam welding, cold forming of hemispherical ends of reentrant cans, and a large variety of rolled joints of innovative designs. This experience provided the foundation for taking up and successfully completing several tasks relating to coolant channels, liquid poison channels and sparger channels for PHWRs and test sections for the in-pile loops of Dhruva reactor. For life prediction and safety assessment of coolant channels of PHWRs some analytical tools, notably, a computer code for prediction of creep limited life of coolant channels has been developed. Some of the future challenges include the development of easily replaceable coolant channels and also large diameter coolant channels for Advanced Heavy Water Reactor, and development of solutions to overcome deterioration of service life of coolant channels due to hydriding. (author). 5 refs., 13 figs., 1 tab

Activities in the Czech Republic for reactor pressure components lifetime management

International Nuclear Information System (INIS)

Brumovsky, M.

1994-01-01

The following activities in the Czech republic for reactor pressure components lifetime management are described: upgrading and safety assurance of nuclear power plants (NPP) with reactors of WWER-440/V-230 type, safety assurance of NPPs with reactors of WWER-440/V-213, lifetime management programme of NPPs with WWER-440/V-213 reactors, preparation of start-up of NPPs with WWER-1000/V-320 reactors, preparation of guides for lifetime as well as defect allowability evaluation in main components of primary and secondary circuits. 3 figs

High flux testing reactor Petten. Replacement of the reactor vessel and connected components. Overall report

International Nuclear Information System (INIS)

Chrysochoides, N.G.; Cundy, M.R.; Von der Hardt, P.; Husmann, K.; Swanenburg de Veye, R.J.; Tas, A.

1985-01-01

The project of replacing the HFR originated in 1974 when results of several research programmes confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report contains the detailed chronology of events concerning preparation and execution of the replacement. After a 14 months' outage the reactor resumed routine operation on 14th February, 1985. At the end of several years of planning and preparation the reconstruction proceded in the following steps: unloading of the old core, decay of short-lived radioactivity in December 1983, removal of the old tank and of its peripheral equipment in January-February 1984, segmentation and waste disposal of the removed components in March-April, decontamination of the pools, bottom penetration overhauling in May-June, installation of the new tank and other new components in July-September, testing and commissioning, including minor modifications in October-December, and, trials runs and start-up preparation in January-February 1985. The new HFR Petten features increased and improved experimental facilities. Among others the obsolete thermal columns was replaced by two high flux beam tubes. Moreover the new plant has been designed for future increases of reactor power and neutron fluxes. For the next three to four years the reactor has to cope with a large irradiation programme, claiming its capacity to nearly 100%

Characteristics of outage radiation fields around various reactor components

International Nuclear Information System (INIS)

Verzilov, Y.; Husain, A.; Corbin, G.

2008-01-01

Full text: Activity monitoring surveys, consisting of gamma spectroscopy and dose rate measurements, of various CANDU station components such as the reactor face, feeder cabinet, steam generators and moderator heat exchangers are often performed during shutdown in order to trend the transport of activity around the primary heat transport and moderator systems. Recently, the increased dose expenditure for work such as feeder inspection and replacement in the reactor vault has also spurred interest in improved characterization of the reactor face fields to facilitate better ALARA decision making and hence a reduction in future dose expenditures. At present, planning for reactor face work is hampered by insufficient understanding of the relative contribution of the various components to the overall dose. In addition to the increased dose expenditure for work at the reactor face, maintenance work associated with horizontal flux detectors and liquid injection systems has also resulted in elevated dose expenditures. For instance at Darlington, radiation fields in the vicinity of horizontal flux detectors (HFD) and Liquid Injection Shutdown System (LISS) nozzle bellows are trending upwards with present contact fields being in the range 16-70 rem/h and working distance fields being in the range 100-500 mrem/h. This paper presents findings based on work currently being funded by the CANDU Owners Group. Measurements were performed at Ontario Power Generation's Pickering and Darlington nuclear stations. Specifically, the following are addressed: Characteristics of Reactor Vault Fields; Characteristics of Steam Generator Fields; Characteristics of Moderator Heat Exchanger Fields. Measurements in the reactor vault were performed at the reactor face, along the length of end fittings, along the length of feeders, at the bleed condenser and at the HFD and LISS nozzle bellows. Steam generator fields were characterized at various elevations above the tube sheet, with and without the

Strategic vehicle fleet management - the make or buy problem

Directory of Open Access Journals (Sweden)

Adam Redmer

2014-06-01

Full Text Available Background: Fleets constitute the most important production means in transportation. Their appropriate management is crucial for all companies having transportation duties. The paper is the first one of a series of three papers that the author dedicates to the strategic vehicle fleet management topic. Methods: The paper discusses ways of fulfilling company's transportation needs (MAKE-or-BUY problem. It means the choice between using company's own and outside fleet (buying transportation services in a market. The essence of the MAKE-or-BUY problem lies in a time dependency, a seasonal nature of transportation needs. It leads to the MAKE-and-BUY solutions including utilization of both in-house and outside fleets. In the paper an original mathematical model (an optimization method allowing for the MAKE-and-BUY analysis is proposed. Results: An application of the proposed optimization method in a real-life decision situation (the case study within the Polish environment and the obtained solution are presented. The solution shows a low economic justification for using the MAKE option in practice. Especially when a fleet composed of brand new vehicles is considered. Conclusions: The paper will be continued in two further papers dedicated to strategic vehicle fleet management problems including fleet sizing / composition and fleet replacement.

Three-dimensional reactor dynamics code for VVER type nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

Three-dimensional reactor dynamics code for VVER type nuclear reactors

International Nuclear Information System (INIS)

Kyrki-Rajamaeki, R.

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

Examination of core components removed from CANDU reactors

International Nuclear Information System (INIS)

Cheadle, B.A.; Coleman, C.E.; Rodgers, D.K.; Davies, P.H.; Chow, C.K.; Griffiths, M.

1988-11-01

Components in the core of a nuclear reactor degrade because the environment is severe. For example, in CANDU reactors the pressure tubes must contend with the effects of hot pressurised water and damage by a flux of fast neutrons. To evaluate any deterioration of components and determine the cause of the occasional failure, we have developed a wide range of remote-handling techniques to examine radioactive materials. As well as pressure tubes, we have examined calandria tubes, garter springs, end fittings, liquid-zone control units and flux detectors. The results from these examinations have produced solutions to problems and continually provide information to help understand the processes that may limit the lifetime of a component

Multi-objective optimization for nuclear fleet evolution scenarios using COSI - 5187

International Nuclear Information System (INIS)

Freynet, D.; Coquelet-Pascal, C.; Eschbach, R.; Krivtchik, G.; Merle-Lucotte, E.

2015-01-01

The consequences of strategic choices on material inventories and flux in the fuel cycle can be analysed with the COSI code. Indeed COSI enables to compare various fleet evolution options (e.g. new reactor systems deployment) and different nuclear materials management (e.g. plutonium multi-recycling). COSI is coupled with the CESAR depletion code. In this paper a methodology for the nuclear fleet evolution scenarios optimization using COSI is introduced. A large number of scenario calculations is needed to solve an optimization problem, which makes infeasible an optimization calculation using the COSI/CESAR version. Given that CESAR calculations represent about 95% of the COSI computation time, CESAR irradiation surrogate models carrying out with ANN regression method and cooling analytic models have been coupled with COSI. An example of optimization study is presented involving 2 discrete variables related to the number of deployed SFR to renew the French PWR fleet and 2 criteria: minimizing the natural uranium consumption and the number of produced HLW vitrified packages

NATIONAL FLEET DEVELOPMENT IN THE INNOVATIVE ECONOMY. CASE STUDY OF THE GREEK FLEET

Directory of Open Access Journals (Sweden)

Marek Grzybowski

2016-06-01

Full Text Available In the article the development of the Greek fleet in the period of last ten years was discussed. The Greek fleet is an example of accommodating itself to the requirements of the global and innovative economy. Greek shipowners are developing their fleets through the consolidation and replacing older ships with new generation vessels. They are invest-ing into ships adapted for new markets, including LNG maritime transport market. As a result of it Greece became a market leader in maritime transport sector and their fleet a biggest and youngest fleet of world in 2016.

Maximizing your fleet

Energy Technology Data Exchange (ETDEWEB)

Lytle, J. [GE Capital Rail Services, Calgary, AB (Canada)

2001-07-01

A series of viewgraphs were presented to illustrate this discussion which focused on the economics of the railroad environment in 2001 and beyond. Fleet productivity and customer relations were described. The viewgraphs were entitled: new car builds; fleet demographics for North American LPG/AA fleet; and, issues and trends of North American LPG/AA fleet. It was noted that there is a continued demand for North American LPG/AA fleet as cars are phased out. GE Capital Rail Services is addressing the future by focusing on the following four initiatives: service, globalization, building a six sigma company, and transforming into an e-business. The company's six sigma approach is based on customer orientation, safety and compliance, quality and reliability, productivity and planning, cost, and people and culture. The actions taken thus far have been a planned maintenance program, the relocation of safety valves, and the elimination of early failures. figs.

FLEET ASSIGNMENT MODELLING

Directory of Open Access Journals (Sweden)

2016-01-01

Full Text Available The article is devoted to the airline scheduling process and methods of its modeling. This article describes the main stages of airline scheduling process (scheduling, fleet assignment, revenue management, operations, their features and interactions. The main part of scheduling process is fleet assignment. The optimal solution of the fleet assignment problem enables airlines to increase their incomes up to 3 % due to quality improving of connections and execution of the planned number of flights operated by less number of aircraft than usual or planned earlier. Fleet assignment of scheduling process is examined and Conventional Leg-Based Fleet Assignment Model is analyzed. Finally strong and weak aspects of the model (SWOT are released and applied. The article gives a critical analysis of FAM model, with the purpose of identi- fying possible options and constraints of its use (for example, in cases of short-term and long-term planning, changing the schedule or replacing the aircraft, as well as possible ways to improve the model.

Centralized Reliability Data Organization (CREDO) assessment of critical component unavailability in liquid metal reactors

International Nuclear Information System (INIS)

Koger, K.H.; Haire, M.J.; Humphrys, B.L.; Manneschmidt, J.F.; Setoguchi, K.; Nakai, R.

1988-01-01

The Centralized Reliability Data Organization (CREDO) is the largest repository of liquid metal reactor (LMR) component reliability data in the world. It is jointly sponsored by the US Dept. of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. The CREDO data base contains information on a population of more than 20,000 components and approximately 1500 event records. A conservative estimation is that the total component operating hours is approaching 2.2 billion hours. The work reported here focuses on the availability information contained in CREDO and the development of availability critical items lists. That is, individual components are ranked in prioritized lists from worst to best performers from an availability standpoint. Availability as used here is an inherent characteristics of the component and is not necessarily related to plant operability. A major observation is that a few components have a much higher unavailability factor than the average. The top fifteen components contribute 93%, 77%, and 87% of the total system unavailability for EBR-II, FFTF, and JOYO respectively. Critical components common to all three sites are mechanical pumps and electromagnetic pumps. Application of resources to these components with the highest unavailability will have the greatest effect on overall availability. All three sites demonstrate that low maintainability (i.e., long repair times), rather than unreliability (i.e., high failure rates), are the main contributors, by about a two-to-one margin, to liquid metal system unavailability

Residual stress improving method for reactor structural component and residual stress improving device therefor

Energy Technology Data Exchange (ETDEWEB)

Enomoto, Kunio; Otaka, Masahiro; Kurosawa, Koichi; Saito, Hideyo; Tsujimura, Hiroshi; Tamai, Yasukata; Urashiro, Keiichi; Mochizuki, Masato

1996-09-03

The present invention is applied to a BWR type reactor, in which a high speed jetting flow incorporating cavities is collided against the surface of reactor structural components to form residual compression stresses on the surface layer of the reactor structural components thereby improving the stresses on the surface. Namely, a water jetting means is inserted into the reactor container filled with reactor water. Purified water is pressurized by a pump and introduced to the water jetting means. The purified water jetted from the water jetting means and entraining cavities is abutted against the surface of the reactor structural components. With such procedures, since the purified water is introduced to the water jetting means by the pump, the pump is free from contamination of radioactive materials. As a result, maintenance and inspection for the pump can be facilitated. Further, since the purified water injection flow entraining cavities is abutted against the surface of the reactor structural components being in contact with reactor water, residual compression stresses are exerted on the surface of the reactor structural components. As a result, occurrence of stress corrosion crackings of reactor structural components is suppressed. (I.S.)

Residual stress improving method for reactor structural component and residual stress improving device therefor

International Nuclear Information System (INIS)

Enomoto, Kunio; Otaka, Masahiro; Kurosawa, Koichi; Saito, Hideyo; Tsujimura, Hiroshi; Tamai, Yasukata; Urashiro, Keiichi; Mochizuki, Masato.

1996-01-01

The present invention is applied to a BWR type reactor, in which a high speed jetting flow incorporating cavities is collided against the surface of reactor structural components to form residual compression stresses on the surface layer of the reactor structural components thereby improving the stresses on the surface. Namely, a water jetting means is inserted into the reactor container filled with reactor water. Purified water is pressurized by a pump and introduced to the water jetting means. The purified water jetted from the water jetting means and entraining cavities is abutted against the surface of the reactor structural components. With such procedures, since the purified water is introduced to the water jetting means by the pump, the pump is free from contamination of radioactive materials. As a result, maintenance and inspection for the pump can be facilitated. Further, since the purified water injection flow entraining cavities is abutted against the surface of the reactor structural components being in contact with reactor water, residual compression stresses are exerted on the surface of the reactor structural components. As a result, occurrence of stress corrosion crackings of reactor structural components is suppressed. (I.S.)

Comparison of three ICF reactor designs

International Nuclear Information System (INIS)

Hogan, W.J.

1984-01-01

Three concepts for inertial confinement fusion (ICF) reactors are described and compared with each other, and with magnetic fusion and fission reactors on the basis of environmental impact, safety and efficiency. The critical technical developments of each concept are described. The three concepts represent alternative development paths for inertial fusion

Design codes for gas cooled reactor components

International Nuclear Information System (INIS)

1990-12-01

High-temperature gas-cooled reactor (HTGR) plants have been under development for about 30 years and experimental and prototype plants have been operated. The main line of development has been electricity generation based on the steam cycle. In addition the potential for high primary coolant temperature has resulted in research and development programmes for advanced applications including the direct cycle gas turbine and process heat applications. In order to compare results of the design techniques of various countries for high temperature reactor components, the IAEA established a Co-ordinated Research Programme (CRP) on Design Codes for Gas-Cooled Reactor Components. The Federal Republic of Germany, Japan, Switzerland and the USSR participated in this Co-ordinated Research Programme. Within the frame of this CRP a benchmark problem was established for the design of the hot steam header of the steam generator of an HTGR for electricity generation. This report presents the results of that effort. The publication also contains 5 reports presented by the participants. A separate abstract was prepared for each of these reports. Refs, figs and tabs

Improving fleet performance

International Nuclear Information System (INIS)

Ramjist, S.

2015-01-01

Use Fleet Initiatives to Improve Overall Fleet Performance . Tightly Integrated with Business Planning (Cause & Effect) . Leverage Strength of Broader Organization - Converge on Standard Business Practices . Ancillary Benefit of Improved Agility.

Improving fleet performance

Energy Technology Data Exchange (ETDEWEB)

Ramjist, S. [Ontario Power Generation, Toronto, ON (Canada)

2015-07-01

Use Fleet Initiatives to Improve Overall Fleet Performance . Tightly Integrated with Business Planning (Cause & Effect) . Leverage Strength of Broader Organization - Converge on Standard Business Practices . Ancillary Benefit of Improved Agility.

AGING MANAGMENT OF REACTOR COOLANT SYSTEM MECHANICAL COMPONENTS FOR LICENSE RENEWAL

International Nuclear Information System (INIS)

SUBUDHI, M.; MORANTE, R.; LEE, A.D.

2002-01-01

The reactor coolant system (RCS) mechanical components that require an aging management review for license renewal include the primary loop piping and associated connections to other support systems, reactor vessel, reactor vessel internals, pressurizer. steam generators, reactor coolant pumps, and all other inter-connected piping, pipe fittings, valves, and bolting. All major RCS components are located inside the reactor building. Based on the evaluation findings of recently submitted license renewal applications for pressurized water reactors, this paper presents the plant programs and/or activities proposed by the applicants to manage the effects of aging. These programs and/or activities provide reasonable assurance that the intended function(s) of these mechanical components will be maintained for the period of extended operation. The license renewal application includes identification of RCS subcomponents that are within the scope of license renewal and are vulnerable to age-related degradation when exposed to environmental and operational conditions. determination of the effects of aging on their intended safety functions. and implementation of the aging management programs and/or activities including both current and new programs. Industry-wide operating experience, including generic communication by the NRC, is part of the aging management review for the RCS components. In addition, this paper discusses time-limited aging analyses associated with neutron embrittlement of the reactor vessel beltline region and thermal fatigue

External vibrations measurement of reactor components

Energy Technology Data Exchange (ETDEWEB)

Rogers, S A [Nuclear Electric plc, Barnwood (United Kingdom); Sugden, J [Magnox Electric, Berkeley (United Kingdom)

1997-12-31

The paper outlines the use of External Vibration Monitoring for remote vibration assessment of internal reactor components. The main features of the technique are illustrated by a detailed examination of the specific application to the problem of Heysham 2 Fuel Plug Unit monitoring. (author). 6 figs.

Chemical decontamination of reactor components

International Nuclear Information System (INIS)

Riess, R.; Berthold, H.O.

1977-08-01

A solution for the decontamination of reactor components of the primary system was developed. This solution is a modification of the APAC- (Alkaline Permanganate Ammonium Citrate) system described in the literature. The most important advantage of the present solution over the APAC-method is that it does not induce any selective corrosion attack on materials like stainless steel (austenitic), Inconel 600 and Incoloy 800. (orig.) [de

Zinc injection on the EDF fleet monitoring the injection on 12 units

International Nuclear Information System (INIS)

Le-Meur, Gaelle Harmand; Anne-Marie; Stutzmann, Agnes; Taunier, Stephane; Benfarah, Moez; Bretelle, Jean-Luc; Alain, Rocher; Claeys, Myriam; Bonne, Sebastien

2012-09-01

After a first implementation of zinc injection at Bugey 2 and Bugey 4, EDF decided to extend the program to other units of its fleet. 14 more reactors from the French fleet of 58 were chosen in order to - Reduce the radiation sources for curative or preventive (after SGR) reasons - Mitigate stress corrosion cracking on nickel alloys and reduce the rate of generalized corrosion - Prevent the risk of CIPS, mainly after a fuel management change. Zinc injection started on 9 new units in 2011, 1 unit in 2012 and will be extended to 4 other units before the end of 2013. To monitor the injection, EDF has defined a complete program concerning chemistry, radiation protection (dose rate and deposited activities measurements), materials (statistical analysis of SG tube cracks), fuel (oxide measurements) and waste (radiochemical characterization of filters). Reference units were chosen for each field because of the size of the fleet. This paper will detail the different monitoring programs on the EDF plants injecting zinc. (authors)

Fleet DNA (Presentation)

Energy Technology Data Exchange (ETDEWEB)

Walkokwicz, K.; Duran, A.

2014-06-01

The Fleet DNA project objectives include capturing and quantifying drive cycle and technology variation for the multitude of medium- and heavy-duty vocations; providing a common data storage warehouse for medium- and heavy-duty vehicle fleet data across DOE activities and laboratories; and integrating existing DOE tools, models, and analyses to provide data-driven decision making capabilities. Fleet DNA advantages include: for Government - providing in-use data for standard drive cycle development, R&D, tech targets, and rule making; for OEMs - real-world usage datasets provide concrete examples of customer use profiles; for fleets - vocational datasets help illustrate how to maximize return on technology investments; for Funding Agencies - ways are revealed to optimize the impact of financial incentive offers; and for researchers -a data source is provided for modeling and simulation.

Transportation capabilities of the existing cask fleet

International Nuclear Information System (INIS)

Johnson, P.E.; Wankerl, M.W.; Joy, D.S.

1991-01-01

This paper describes a number of scenarios estimating the amount of spent nuclear fuel that could be transported to a Monitored Retrievable Storage (MRS) Facility by various combinations of existing cask fleets. To develop the scenarios, the data provided by the Transportation System Data Base (TSDB) were modified to reflect the additional time for cask turnaround resulting from various startup and transportation issues. With these more realistic speed and cask-handling assumptions, the annual transportation capability of a fleet consisting of all of the existing casks is approximately 465 metric tons of uranium (MTU). The most likely fleet of existing casks that would be made available to the DOE consists of two rail, three overweight truck, and six legal weight truck casks. Under the same transportation assumptions, this cask fleet is capable of approximately transporting 270 MTU/year. These ranges of capability is a result of the assumptions pertaining to the number of casks assumed to be available. It should be noted that this assessment assumes additional casks based on existing certifications are not fabricated

Transportation capabilities of the existing cask fleet

International Nuclear Information System (INIS)

Johnson, P.E.; Joy, D.S.; Wankerl, M.W.

1991-01-01

This paper describes a number of scenarios estimating the amount of spent nuclear fuel that could be transported to a Monitored Retrievable Storage (MRS) Facility by various combinations of existing cask fleets. To develop the scenarios, the data provided by the Transportation System Data Base (TSDB) were modified to reflect the additional time for cask turnaround resulting from various startup and transportation issues. With these more realistic speed and cask-handling assumptions, the annual transportation capability of a fleet consisting of all of the existing casks is approximately 46 metric tons of uranium (MTU). The most likely fleet of existing casks that would be made available to the Department of Energy (DOE) consists of two rail, three overweight truck, and six legal weight truck casks. Under the same transportation assumptions, this cask fleet is capable of approximately transporting 270 MTU/year. These ranges of capability is a result of the assumptions pertaining to the number of casks assumed to be available. It should be noted that this assessment assumes additional casks based on existing certifications are not fabricated. 5 refs., 4 tabs

Scale modeling flow-induced vibrations of reactor components

International Nuclear Information System (INIS)

Mulcahy, T.M.

1982-06-01

Similitude relationships currently employed in the design of flow-induced vibration scale-model tests of nuclear reactor components are reviewed. Emphasis is given to understanding the origins of the similitude parameters as a basis for discussion of the inevitable distortions which occur in design verification testing of entire reactor systems and in feature testing of individual component designs for the existence of detrimental flow-induced vibration mechanisms. Distortions of similitude parameters made in current test practice are enumerated and selected example tests are described. Also, limitations in the use of specific distortions in model designs are evaluated based on the current understanding of flow-induced vibration mechanisms and structural response

Sodium removal from Hallam Reactor components

International Nuclear Information System (INIS)

Huntsman, L.K.; Meservey, R.H.

1979-08-01

This report discussed the removal of sodium from major components of the Hallam Nuclear Power Facility. This facility contained the experimental ractor used to test the feasibility of sodium coolant. The Idaho Operations Office of the Department of Energy assigned EG and G Idaho, Inc., the task of carrying out this decontamination and decommissioning program at the Idaho National Engineering Laboratory (INEL). Since their shipment to the INEL from Lincoln, Nebraska in 1968, the Hallam Reactor components had been stored in inert nitrogen to prevent the sodium in the components from reacting with moisture in the air. The procedure used to react the sodium in the components and to decontaminate them is discussed. Problems and unusual occurrences in the decontamination and decommissioning process are also reported

Data base formation for important components of reactor TRIGA MARK II

International Nuclear Information System (INIS)

Jordan, R.; Mavko, B.; Kozuh, M.

1992-01-01

The paper represents specific data base formation for reactor TRIGA MARK II in Podgorica. Reactor operation data from year 1985 to 1990 were collected. Two groups of collected data were formed. The first group includes components data and the second group covers data of reactor scrams. Time related and demand related models were used for data evaluation. Parameters were estimated by classical method. Similar data bases are useful everywhere where components unavailabilities may have severe drawback. (author) [sl

How to bring the nuclear share to 50 pc without stopping any reactor, nor increasing CO2 emissions. The issue of renewal of the French nuclear reactor fleet is also part of the debate

International Nuclear Information System (INIS)

Nifenecker, Herve

2013-01-01

In a first article, after having recalled the present situation of the French energy mix, the author analyses the consequences of the shutting down of 22 nuclear reactors by 2025 to bring the nuclear share down from 75 to 50 per cent. Different options are discussed: fossil energy (but CO 2 emissions would be impacted), renewable energy (hydroelectricity cannot be further developed, biomass can be useful for heat networks, and wind energy is limited). The author examines the possibilities of decreasing electricity consumption, and of a significant contribution of gas plants. He compares direct gas- and fuel oil-based heating with heating with electricity produced by gas plants. A second article addresses the issue of renewal of the French nuclear reactor fleet. It discusses how to choose the right renewal pace, and an assessment of exploitation losses

Travelling cranes for heavy reactor component handling

International Nuclear Information System (INIS)

Champeil, M.

1977-01-01

Structure and operating machinery of two travelling cranes (600 t and 450 t) used in the Framatome factory for handling heavy reactor components are described. When coupled, these cranes can lift loads up to 1000 t [fr

RCC-MRx: Design and construction rules for mechanical components in high-temperature structures, experimental reactors and fusion reactors

International Nuclear Information System (INIS)

2015-01-01

The RCC-MRx code was developed for sodium-cooled fast reactors (SFR), research reactors (RR) and fusion reactors (FR-ITER). It provides the rules for designing and building mechanical components involved in areas subject to significant creep and/or significant irradiation. In particular, it incorporates an extensive range of materials (aluminum and zirconium alloys in response to the need for transparency to neutrons), sizing rules for thin shells and box structures, and new modern welding processes: electron beam, laser beam, diffusion and brazing. The RCC-MR code was used to design and build the prototype Fast Breeder Reactor (PFBR) developed by IGCAR in India and the ITER Vacuum Vessel. The RCC-Mx code is being used in the current construction of the RJH experimental reactor (Jules Horowitz reactor). The RCC-MRx code is serving as a reference for the design of the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration), for the design of the primary circuit in MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) and the design of the target station of the ESS project (European Spallation Source). Contents of the 2015 edition of the RCC-MRx code: Section I General provisions; Section II Additional requirements and special provisions; Section III Rules for nuclear installation mechanical components: Volume I: Design and construction rules: Volume A (RA): General provisions and entrance keys, Volume B (RB): Class 1 components and supports, Volume C (RC): Class 2 components and supports, Volume D (RD): Class 3 components and supports, Volume K (RK): Examination, handling or drive mechanisms, Volume L (RL): Irradiation devices, Volume Z (Ai): Technical appendices; Volume II: Materials; Volume III: Examinations methods; Volume IV: Welding; Volume V: Manufacturing operations; Volume VI: Probationary phase rules

Hydrodynamic impact of reactor components - a review

International Nuclear Information System (INIS)

Krajcinovic, D.

1977-01-01

A variety of components belonging to a nuclear reactor are by virtue of their design exposed to a mass of fluid which is either in motion or can be set into motion under certain conditions. While the reactor is in its operational mode, the excitations of the structure by the fluid are generally of moderate intensities. In the case of a well designed component, these pressure fluctuations should not cause the failure of the structure. Problems of this type, generally known as vibrations of structures immersed into fluid (under either periodic or random excitations) have been studied in the past rather extensively. In an upset or emergency condition, a pressure pulse is usually generated and propagated through the fluid. While this hypothetical event is an occurrence of low probability the associated pressures are, as a rule, of intensities sufficiently large to cause extensive damage or even the failure of the component. This type of transient interaction problem is much less studied and the aim of this review is to offer a brief discussion of some of the more interesting results. (Auth.)

Fleet operator risks for using fleets for V2G regulation

International Nuclear Information System (INIS)

Hill, Davion M.; Agarwal, Arun S.; Ayello, Francois

2012-01-01

Future fleets of vehicles may include electric vehicles (EVs) or hybrid electric vehicles (HEVs) because of potential fuel savings. Recent demonstration of diesel parallel hybrids in a delivery fleet led to fuel economy improvements, and hybrid bus demonstrations exhibited twice the fuel economy of the conventional bus. Fleet ownership may include management of a fleet of vehicles as small as 10 units and as large as hundreds or thousands. In addition to fuel savings, the newer extended range electric vehicles (EREVs) and pure EVs permit vehicle to grid (V2G) opportunities. These V2G opportunities may present additional revenue for fleets by providing ancillary services to local grid independent system operators (ISOs), provided that the burden of driving and V2G services do not accelerate the degradation of the battery systems in these vehicles. The subject of this study is to determine the financial risks associated with accelerated battery degradation in a V2G-enabled EREV fleet expected to perform ancillary service duty while charging in addition to the normal loads of drive cycle duty. We determine that battery cycle life during V2G duty is a critical parameter, which can determine whether or not the business model is viable. - Highlights: ► V2G regulation cycle life of EREV batteries must be >50,000 cycles to be profitable. ► Present knowledge and test data about the impact of V2G cycles on battery life are limited. ► Replacement of batteries and energy throughput are major factors in the cost–benefit analysis. ► V2G fleet is not viable with present data but can be viable with some technical advancement.

The Thermal-hydraulic Analysis for the Aging Effect of the Component in CANDU-6 Reactor

International Nuclear Information System (INIS)

Bae, Jun Ho; Jung, Jong Yeob

2014-01-01

CANDU reactor consists of a lot of components, including pressure tube, reactor pump, steam generator, feeder pipe, and so on. These components become to have the aging characteristics as the reactor operates for a long time. The aging phenomena of these components lead to the change of operating parameters, and it finally results to the decrease of the operating safety margin. Actually, due to the aging characteristics of components, CANDU reactor power plant has the operating license for the duration of 30 years and the plant regularly check the plant operating state in the overhaul period. As the reactor experiences the aging, the reactor operators should reduce the reactor power level in order to keep the minimum safety margin, and it results to the deficit of economical profit. Therefore, in order to establish the safety margin for the aged reactor, the aging characteristics for components should be analyzed and the effect of aging of components on the operating parameter should be studied. In this study, the aging characteristics of components are analyzed and revealed how the aging of components affects to the operating parameter by using NUCIRC code. Finally, by scrutinizing the effect of operating parameter on the operating safety margin, the effect of aging of components on the safety margin has been revealed

Replacement of core components in the Advanced Test Reactor

International Nuclear Information System (INIS)

Durney, J.L.; Croucher, D.W.

1990-01-01

The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. The methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and is expected to continue operation for at least and additional 25 years. Aging evaluations are in progress to address additional replacements that may be needed during this period

Replacement of core components in the Advanced Test Reactor

International Nuclear Information System (INIS)

Durney, J.L.; Croucher, D.W.

1989-01-01

The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. Methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and will continue operation for perhaps another 20 years. Aging evaluations are in program to address additional replacements that may be needed during this extended time period. 3 figs

Thermal-hydraulic limitations on water-cooled fusion reactor components

International Nuclear Information System (INIS)

Cha, Y.S.; Misra, B.

1986-01-01

An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on conventional design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require unconventional shapes or heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for complex geometries, especially under asymmetrically heated conditions. Since the data presented herein are concerned primarily with thermal protection of the reactor components, the final design should consider other factors such as thermal stresses, temperature limits, and fatigue

TEMAS: fleet-based bio-economic simulation software to evaluate management strategies accounting for fleet behaviour

DEFF Research Database (Denmark)

Ulrich, Clara; Andersen, Bo Sølgaard; Sparre, Per Johan

2007-01-01

TEMAS (technical management measures) is a fleet-based bio-economic software for evaluating management strategies accounting for technical measures and fleet behaviour. It focuses on mixed fisheries in which several fleets can choose among several fishing activities to target different stocks...

National Clean Fleets Partnership (Fact Sheet)

Energy Technology Data Exchange (ETDEWEB)

2012-01-01

Provides an overview of Clean Cities National Clean Fleets Partnership (NCFP). The NCFP is open to large private-sector companies that have fleet operations in multiple states. Companies that join the partnership receive customized assistance to reduce petroleum use through increased efficiency and use of alternative fuels. This initiative provides fleets with specialized resources, expertise, and support to successfully incorporate alternative fuels and fuel-saving measures into their operations. The National Clean Fleets Partnership builds on the established success of DOE's Clean Cities program, which reduces petroleum consumption at the community level through a nationwide network of coalitions that work with local stakeholders. Developed with input from fleet managers, industry representatives, and Clean Cities coordinators, the National Clean Fleets Partnership goes one step further by working with large private-sector fleets.

An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

Energy Technology Data Exchange (ETDEWEB)

Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

2009-11-01

This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

What is behind fleet evolution: a framework for flow analysis and application to the French Atlantic fleet

OpenAIRE

Quillérou, Emmanuelle; Guyader, Olivier

2012-01-01

The study of fishery dynamics considers national-level fleet evolution. It has, however, failed to consider the flows behind fleet evolution as well as the impact of the dynamics of owners of invested capital on fleet evolution. This paper establishes a general conceptual framework which identifies different vessel and owner flows behind fleet evolution and some relationships between these flows. This descriptive conceptual framework aims to change the current focus on drivers of fleet evolut...

Analysis of a Spanish energy scenario with Generation IV nuclear reactors

International Nuclear Information System (INIS)

Ochoa, Raquel; Jimenez, Gonzalo; Perez-Martin, Sara

2013-01-01

Highlights: • Spanish energy scenario for the hypothetical deployment of Gen-IV SFR reactors. • Availability of national resources is assessed, considering SFR’s breeding. • An assessment of the impact of transmuting MA on the final repository. • SERPENT code with own pre- and post-processing tools were employed. • The employed SFR core design is based on the specifications of the CP-ESFR. - Abstract: The advantages of fast-spectrum reactors consist not only of an efficient use of fuel through the breeding of fissile material and the use of natural or depleted uranium, but also of the potential reduction of the amount of actinides such as americium and neptunium contained in the irradiated fuel. The first aspect means a guaranteed future nuclear fuel supply. The second fact is key for high-level radioactive waste management, because these elements are the main responsible for the radioactivity of the irradiated fuel in the long term. The present study aims to analyze the hypothetical deployment of a Gen-IV Sodium Fast Reactor (SFR) fleet in Spain. A nuclear fleet of fast reactors would enable a fuel cycle strategy different than the open cycle, currently adopted by most of the countries with nuclear power. A transition from the current Gen-II to Gen-IV fleet is envisaged through an intermediate deployment of Gen-III reactors. Fuel reprocessing from the Gen-II and Gen-III Light Water Reactors (LWR) has been considered. In the so-called advanced fuel cycle, the reprocessed fuel used to produce energy will breed new fissile fuel and transmute minor actinides at the same time. A reference case scenario has been postulated and further sensitivity studies have been performed to analyze the impact of the different parameters on the required reactor fleet. The potential capability of Spain to supply the required fleet for the reference scenario using national resources has been verified. Finally, some consequences on irradiated final fuel inventory are assessed

A plan for safety and integrity of research reactor components

International Nuclear Information System (INIS)

Moatty, Mona S. Abdel; Khattab, M.S.

2013-01-01

Highlights: ► A plan for in-service inspection of research reactor components is put. ► Section XI of the ASME Code requirements is applied. ► Components subjected to inspection and their classes are defined. ► Flaw evaluation and its acceptance–rejection criteria are reviewed. ► A plan of repair or replacement is prepared. -- Abstract: Safety and integrity of a research reactor that has been operated over 40 years requires frequent and thorough inspection of all the safety-related components of the facility. The need of increasing the safety is the need of improving the reliability of its systems. Diligent and extensive planning of in-service inspection (ISI) of all reactor components has been imposed for satisfying the most stringent safety requirements. The Safeguards Officer's responsibilities of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code ASME Code have been applied. These represent the most extensive and time-consuming part of ISI program, and identify the components subjected to inspection and testing, methods of component classification, inspection and testing techniques, acceptance/rejection criteria, and the responsibilities. The paper focuses on ISI planning requirements for welded systems such as vessels, piping, valve bodies, pump casings, and control rod-housing parts. The weld in integral attachments for piping, pumps, and valves are considered too. These are taken in consideration of safety class (1, 2, 3, etc.), reactor age, and weld type. The parts involve in the frequency of inspection, the examination requirements for each inspection, the examination method are included. Moreover the flaw evaluation, the plan of repair or replacement, and the qualification of nondestructive examination personnel are considered

Stalin's Big Fleet Program

National Research Council Canada - National Science Library

Mauner, Milan

2002-01-01

Although Dr. Milan Hauner's study 'Stalin's Big Fleet program' has focused primarily on the formation of Big Fleets during the Tsarist and Soviet periods of Russia's naval history, there are important lessons...

Safety and radiation protection within the French electronuclear fleet in 2014. The IRSN's opinion

International Nuclear Information System (INIS)

2015-11-01

This annual report concerns the 58 pressurized water reactors of the French electronuclear fleet which are currently being operated. After a presentation of this fleet, and a discussion of the global assessment of safety and radiation protection within this fleet (discussion of the noticed trends), this document first reports the analysis performed by the IRSN of some important events and anomalies which occurred in 2014: the presence of migrating elements in the primary circuit which resulted in the damage fuel assembly sheaths (in Saint-Laurent des Eaux), the spraying of electric equipment due to a water flow from leaky hoppers, and the risk of unavailability of important equipment due to a high temperature in the room of the emergency back-up turbo generator set. The next chapter comments significant evolutions: renewal of the command-control system for the 1300 MWe reactors, risks induced on the Gravelines nuclear plant site by the exploitation of the Dunkirk methane terminal, and internal EDF arbitrages related to declarations regarding events which occurred in Val de Loire nuclear plants. These evolutions illustrate how expertises performed by the IRSN on request by the ASN complement EDF expertises and are a contribution to the improvement of installation safety

Aging assessment and mitigation for major LWR [light water reactor] components

International Nuclear Information System (INIS)

Shah, Y.N.; Ware, A.G.; Conley, D.A.; MacDonald, P.E.; Burns, J.J. Jr.

1989-01-01

This paper summarizes some of the results of the Aging Assessment and Mitigation Project sponsored by the US Nuclear Regulatory Commission (USNRC), Office of Nuclear Regulatory Research. The objective of the project is to develop an understanding of the aging degradation of the major light water reactor (LWR) structures and components and to develop methods for predicting the useful life of these components so that the impact of aging on the safe operation of nuclear power plants can be evaluated and addressed. The research effort consists of integrating, evaluating, and updating the available aging-related information. This paper discusses current accomplishments and summarizes the significant degradation processes active in two major components: pressurized water reactor pressurizer surge and spray lines and nozzles, and light water reactor primary coolant pumps. This paper also evaluates the effectiveness of the current inservice inspection programs and presents conclusions and recommendations related to aging of these two major components. 37 refs., 7 figs., 3 tabs

On the classification of structures, systems and components of nuclear research and test reactors

International Nuclear Information System (INIS)

Mattar Neto, Miguel

2009-01-01

The classification of structures, systems and components of nuclear reactors is a relevant issue related to their design because it is directly associated with their safety functions. There is an important statement regarding quality standards and records that says Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. The definition of the codes, standards and technical requirements applied to the nuclear reactor design, fabrication, inspection and tests may be seen as the main result from this statement. There are well established guides to classify structures, systems and components for nuclear power reactors such as the Pressurized Water Reactors but one can not say the same for nuclear research and test reactors. The nuclear reactors safety functions are those required to the safe reactor operation, the safe reactor shutdown and continued safe conditions, the response to anticipated transients, the response to potential accidents and the control of radioactive material. So, it is proposed in this paper an approach to develop the classification of structures, systems and components of these reactors based on their intended safety functions in order to define the applicable set of codes, standards and technical requirements. (author)

State and Alternative Fuel Provider Fleets - Fleet Compliance Annual Report: Model Year 2015, Fiscal Year 2016

Energy Technology Data Exchange (ETDEWEB)

2016-12-01

The U.S. Department of Energy (DOE) regulates covered state government and alternative fuel provider fleets, pursuant to the Energy Policy Act of 1992 (EPAct), as amended. Covered fleets may meet their EPAct requirements through one of two compliance methods: Standard Compliance or Alternative Compliance. For model year (MY) 2015, the compliance rate with this program for the more than 3011 reporting fleets was 100%. More than 294 fleets used Standard Compliance and exceeded their aggregate MY 2015 acquisition requirements by 8% through acquisitions alone. The seven covered fleets that used Alternative Compliance exceeded their aggregate MY 2015 petroleum use reduction requirements by 46%.

Safety classification of systems, structures, and components for pool-type research reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Ryong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-08-15

Structures, systems, and components (SSCs) important to safety of nuclear facilities shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions. Although SSC classification guidelines for nuclear power plants have been well established and applied, those for research reactors have been only recently established by the International Atomic Energy Agency (IAEA). Korea has operated a pool-type research reactor (the High Flux Advanced Neutron Application Reactor) and has recently exported another pool-type reactor (Jordan Research and Training Reactor), which is being built in Jordan. Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance. Because the SSCs in pool-type research reactors are not the pressure-retaining components, codes and standards for design of the SSCs following the safety classification can be selected in a graded approach.

Modular core component support for nuclear reactor

International Nuclear Information System (INIS)

Finch, L.M.; Anthony, A.J.

1975-01-01

The core of a nuclear reactor is made up of a plurality of support modules for containing components such as fuel elements, reflectors and control rods. Each module includes a component support portion located above a grid plate in a low-pressure coolant zone and a coolant inlet portion disposed within a module receptacle which depends from the grid plate into a zone of high-pressure coolant. Coolant enters the module through aligned openings within the receptacle and module inlet portion and flows upward into contact with the core components. The modules are hydraulically balanced within the receptacles to prevent expulsion by the upward coolant forces. (U.S.)

An Assessment of Remote Visual Methods to Detect Cracking in Reactor Components

International Nuclear Information System (INIS)

Cumblidge, Stephen E.; Anderson, Michael T.; Doctor, Steven R.; Simonen, Fredric A.; Elliot, Anthony J.

2008-01-01

Recently, the U.S. nuclear industry has proposed replacing current volumetric and/or surface examinations of certain components in commercial nuclear power plants, as required by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Inservice Inspection of Nuclear Power Plant Components, with a simpler visual testing (VT) method. The advantages of VT are that these tests generally involve much less radiation exposure and time to perform the examination than do volumetric examinations such as ultrasonic testing. The issues relative to the reliability of VT in determining the structural integrity of reactor components were examined. Some piping and pressure vessel components in a nuclear power station are examined using VT as they are either in high radiation fields or component geometry precludes the use of ultrasonic testing (UT) methodology. Remote VT with radiation-hardened video systems has been used by nuclear utilities to find cracks in pressure vessel cladding in pressurized water reactors, core shrouds in boiling water reactors, and to investigate leaks in piping and reactor components. These visual tests are performed using a wide variety of procedures and equipment. The techniques for remote VT use submersible closed-circuit video cameras to examine reactor components and welds. PNNL conducted a parametric study that examined the important variables influencing the effectiveness of a remote visual test. Tested variables included lighting techniques, camera resolution, camera movement, and magnification. PNNL also conducted a limited laboratory test using a commercial visual testing camera system to experimentally determine the ability of the camera system to detect cracks of various widths under ideal conditions. The results of these studies and their implications are presented in this paper

UK fast reactor components. Sodium removal decontamination and requalification

International Nuclear Information System (INIS)

Donaldson, D.M.; Bray, J.A.; Newson, I.H.

1978-01-01

Extensive experience gained at the U.K.A.E.A. Dounreay Nuclear Power Development Establishment is being applied to form the basis of the plant to be provided for sodium removal, decontamination, and requalification of components in future commercial fast reactors. In the first part of a three part paper, the factors to be taken into account, showing the UK philosophy and approach to maintenance and repair operations are discussed. In the second part, PFR facilities for sodium removal and decontamination are described and some examples are given of cleaning components such as pumps, charge machine, cold trap baskets, and steam generator units. Similar facilities at DFR are briefly described. In the third part of the paper a short description is given of the Harwell mass transfer loop, currently used to study the deposition of activated stainless steel corrosion products. Decontamination method for pipework specimens cut from the loop are described and results of first screening tests of various chemical decontaminants are presented. (U.K.)

Reactor materials program process water component failure probability

International Nuclear Information System (INIS)

Daugherty, W. L.

1988-01-01

The maximum rate loss of coolant accident for the Savannah River Production Reactors is presently specified as the abrupt double-ended guillotine break (DEGB) of a large process water pipe. This accident is not considered credible in light of the low applied stresses and the inherent ductility of the piping materials. The Reactor Materials Program was initiated to provide the technical basis for an alternate, credible maximum rate LOCA. The major thrust of this program is to develop an alternate worst case accident scenario by deterministic means. In addition, the probability of a DEGB is also being determined; to show that in addition to being mechanistically incredible, it is also highly improbable. The probability of a DEGB of the process water piping is evaluated in two parts: failure by direct means, and indirectly-induced failure. These two areas have been discussed in other reports. In addition, the frequency of a large bread (equivalent to a DEGB) in other process water system components is assessed. This report reviews the large break frequency for each component as well as the overall large break frequency for the reactor system

Studies on components for a molten salt reactor

International Nuclear Information System (INIS)

Nejedly, M.; Matal, O.

2003-01-01

The aim is contribute to a design of selected components of molten salt reactors with fuel in the molten fluoride salt matrix. Molten salt reactors (MSRs) permit the utilization of plutonium and minor actinides as new nuclear fuel from a traditional nuclear power station with production of electric energy. Results of preliminary feasibility studies of an intermediate heat exchanger, a small power molten salt pump and a modular conception of a steam generator for a demonstration unit of the MSR (30 MW) are summarized. (author)

Tritium in fusion reactor components

International Nuclear Information System (INIS)

Watson, J.S.; Fisher, P.W.; Talbot, J.B.

1980-01-01

When tritium is used in a fusion energy experiment or reactor, several implications affect and usually restrict the design and operation of the system and involve questions of containment, inventory, and radiation damage. Containment is expected to be particularly important both for high-temperature components and for those components that are prone to require frequent maintenance. Inventory is currently of major significance in cases where safety and environmental considerations limit the experiments to very low levels of tritium. Fewer inventory restrictions are expected as fusion experiments are placed in more-remote locations and as the fusion community gains experience with the use of tritium. However, the advent of power-producing experiments with high-duty cycle will again lead to serious difficulties based principally on tritium availability; cyclic operations with significant regeneration times are the principal problems

Extension of the nuclear power plant lifetime

International Nuclear Information System (INIS)

Keramsi, Alain

2011-01-01

After a presentation of the French nuclear context (history of the reactor fleet, choice of reactor type, PWR operation principle, competitiveness, environmental performance), this Power Point presentation addresses the context and challenges of the operation lifetime (average fleet age in different countries, examples of extensions, case of the United States, what is at stake with lifetime extension, decennial visits, EDF strategy), discusses the EDF's safety objectives (definition of the three main safety functions, impact of the operation duration and of the coexistence of two generations for the safety functions), discusses how to manage the ageing phenomenon for replaceable and non-replaceable components

National Clean Fleets Partnership (Fact Sheet)

Energy Technology Data Exchange (ETDEWEB)

None, None

2014-01-01

Clean Cities' National Clean Fleets Partnership establishes strategic alliances with large fleets to help them explore and adopt alternative fuels and fuel economy measures to cut petroleum use. The initiative leverages the strength of nearly 100 Clean Cities coalitions, nearly 18,000 stakeholders, and more than 20 years of experience. It provides fleets with top-level support, technical assistance, robust tools and resources, and public acknowledgement to help meet and celebrate fleets' petroleum-use reductions.

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

Liner Shipping Fleet Repositioning

DEFF Research Database (Denmark)

Tierney, Kevin; Jensen, Rune Møller

2011-01-01

Liner shipping fleet repositioning consists of moving vessels between services in a liner ship- ping network in order to better orient the overall network to the world economy, and to ensure the proper maintenance of vessels. Thus, fleet repositioning involves sailing and loading activities subject...

Thermal aging of some decommissioned reactor components and methodology for life prediction

International Nuclear Information System (INIS)

Chung, H.M.

1989-03-01

Since a realistic aging of cast stainless steel components for end-of-life or life-extension conditions cannot be produced, it is customary to simulate the thermal aging embrittlement by accelerated aging at ∼400 degree C. In this investigation, field components obtained from decommissioned reactors have been examined after service up to 22 yr to provide a benchmark of the laboratory simulation. The primary and secondary aging processes were found to be identical to those of the laboratory-aged specimens, and the kinetic characteristics were also similar. The extent of the aging embrittlement processes and other key factors that are known to influence the embrittlement kinetics have been compared for the decommissioned reactor components and materials aged under accelerated conditions. On the basis of the study, a mechanistic understanding of the causes of the complex behavior in kinetics and activation energy of aging (i.e., the temperature dependence of aging embrittlement between the accelerated and reactor-operating conditions) is presented. A mechanistic correlation developed thereon is compared with a number of available empirical correlations to provide an insight for development of a better methodology of life prediction of the reactor components. 18 refs., 18 figs., 5 tabs

Evaluation of flow-induced vibration prediction techniques for in-reactor components

International Nuclear Information System (INIS)

Mulcahy, T.M.; Turula, P.

1975-05-01

Selected in-reactor components of a hydraulic and structural dynamic scale model of the U. S. Energy Research and Development Administration experimental Fast Test Reactor have been studied in an effort to develop and evaluate techniques for predicting vibration behavior of elastic structures exposed to a moving fluid. Existing analysis methods are used to compute the natural frequencies and modal shapes of submerged beam and shell type components. Component response is calculated, assuming as fluid forcing mechanisms both vortex shedding and random excitations characterized by the available hydraulic data. The free and force vibration response predictions are compared with extensive model flow and shaker test data. (U.S.)

High-temperature-structural design and research and development for reactor system components

International Nuclear Information System (INIS)

Matsumura, Makoto; Hada, Mikio

1985-01-01

The design of reactor system components requires high-temperature-structural design guide with the consideration of the creep effect of materials related to research and development on structural design. The high-temperature-structural design guideline for the fast prototype reactor MONJU has been developed under the active leadership by Power Reactor and Nuclear Fuel Development Corporation and Toshiba has actively participated to this work with responsibility on in-vessel components, performing research and development programs. This paper reports the current status of high-temperature-structural-design-oriented research and development programs and development of analytical system including stress-evaluation program. (author)

Structural integrity of water reactor pressure boundary components

International Nuclear Information System (INIS)

Loss, F.J.

1977-01-01

The dynamic fracture toughness was determined as a function of temperature for three-point bend specimens of A533-B, A508-2, and A302-B steels. Crack propagation rates at 288 0 C in a water reactor environment were determined for A533-B and A508-2. Radiation-induced degradation of notch toughness of reactor steels and welds was explored. The ''warm prestress'' occurring in a flawed reactor vessel following a LOCA and operation of ECCS was studied. 25 figures

A road safety performance indicator for vehicle fleet compatibility.

Science.gov (United States)

Christoph, Michiel; Vis, Martijn Alexander; Rackliff, Lucy; Stipdonk, Henk

2013-11-01

This paper discusses the development and the application of a safety performance indicator which measures the intrinsic safety of a country's vehicle fleet related to fleet composition. The indicator takes into account both the 'relative severity' of individual collisions between different vehicle types, and the share of those vehicle types within a country's fleet. The relative severity is a measure for the personal damage that can be expected from a collision between two vehicles of any type, relative to that of a collision between passenger cars. It is shown how this number can be calculated using vehicle mass only. A sensitivity analysis is performed to study the dependence of the indicator on parameter values and basic assumptions made. The indicator is easy to apply and satisfies the requirements for appropriate safety performance indicators. It was developed in such a way that it specifically scores the intrinsic safety of a fleet due to its composition, without being influenced by other factors, like helmet wearing. For the sake of simplicity, and since the required data is available throughout Europe, the indicator was applied to the relative share of three of the main vehicle types: passenger cars, heavy goods vehicles and motorcycles. Using the vehicle fleet data from 13EU Member States and Norway, the indicator was used to rank the countries' safety performance. The UK was found to perform best in terms of its fleet composition (value is 1.07), while Greece has the worst performance with the highest indicator value (1.41). Copyright © 2013 Elsevier Ltd. All rights reserved.

A development of three-dimensional seismic isolation for advanced reactor systems in Japan: Pt.2

International Nuclear Information System (INIS)

Kenji Takahashi; Kazuhiko Inoue; Asao Kato; Masaki Morishita; Takafumi Fujita

2005-01-01

Two types of three-dimensional seismic isolation systems were developed for the fast breeder reactor (FBR). One is the three-dimensional entire building base isolation system It was developed by collecting concepts Japanese companies from which a combination system with air springs and hydraulic rocking suppression devices was selected. The other is the vertically isolated system for main components with horizontally entire building base isolation, which was developed by adopting coned disk spring devices. In the study, seismic condition was assumed based on a strict reference ground motion. Design data of the building and components are referred to FBR being developed as the 'Commercialized Fast Reactor Cycle System'. Analysis based on these assumed conditions showed suitable combinations of natural frequencies and damping ratios for isolation. Devices were developed to satisfy the combinations. In five years research and development, several verification tests were performed including shake table tests with scaled models. Finally it is found that the two types of seismic isolation systems are available for FBR. The result is reflected in the preliminary design guideline for the three-dimensional isolation system. (authors)

Pressure-tube reactors as a part of Russian nuclear fleet

International Nuclear Information System (INIS)

Gmyrko, V.E.; Grozdov, I.I.; Nikitin, Yu.M.; Petrov, A.A.; Potapov, A.A.; Finyakin, A.F.

2007-01-01

The place and role of channel reactors in nuclear power in our country and the main measures for upgrading and improving the power generating units of nuclear power plants with RBMK reactors are described. It is shown that the risk indicators for serious damage to the core of power generating units with RBMK reactors are lower after upgrading and the corresponding IAEA criterion established for operating nuclear power plants. Upgrading and implementation of a service life extension program has made it possible to obtain licenses for continuing operation of power generating units with first-generation RBMK reactors and predicting a service life increase to 45 years. The characteristics of nuclear power plants with channel reactors with more highly developed internal and natural safety properties are shown in evolutionary designs of the power generating units MCER-860,-1000, and-1500, which have protective shells and which meet all requirements for power generating units built today. It is shown that innovative solutions for the channel reactor concept can be implemented on the basis of the designs of power generating units with nuclear superheating of steam or on the basis of designs for developing reactors with supercritical parameters [ru

The design and implementation of a garbage truck fleet management system

Directory of Open Access Journals (Sweden)

Chen, C. H.

2016-05-01

Full Text Available In recent years, the improvement of cloud computing and mobile computing techniques has led to the availability of a variety of mobile applications (‘apps’ in the app store. For instance, a garbage truck app that can provide the immediate location of a garbage truck, the location of collection points, and forecasted arrival times of garbage trucks would be useful for mobile users. Since the power consumption of apps on mobile devices if of concern to mobile users, an optimised power-saving mechanism for updating messages, which is based on location information, for a proposed garbage truck fleet management system (GTFMS is proposed and implemented in this paper. The GTFMS is a three- component system that includes the on-board units on garbage trucks, a fleet management system, and a garbage truck app. In this study, an arrival time forecasting method is designed and implemented in the fleet management system, so that the garbage truck app can retrieve the forecasted arrival time via web services. A message updating event is then triggered that reports the location of garbage truck and the forecasted arrival time. In experiments conducted on case studies, the results showed that the mean accuracy of predicted arrival time by the proposed method is about 81.45 per cent. As for power consumption, the cost of traditional mobile apps is 2,880 times that of the mechanism proposed in this study. Consequently, the GTFMS can provide the precise forecasted arrival time of garbage trucks to mobile users, while consuming less power.

Fleet Feedback and Fleet Efficiency Metrics

Energy Technology Data Exchange (ETDEWEB)

Singer, Mark R [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

2017-09-05

The Marine Corps have 10 years of experience implementing a telematics program and several lessons to share with partner agencies. This presentation details results of a Marine Corps survey as well as methods of using telematics to promote fleet efficiency and optimize the vehicle acquisition process.

U.S. NRC training for research and training reactor inspectors

International Nuclear Information System (INIS)

Sandquist, G.M.; Kunze, J.F.

2011-01-01

Currently, a large number of license activities (Early Site Permits, Combined Operating License, reactor certifications, etc.), are pending for review before the United States Nuclear Regulatory Commission (US NRC). Much of the senior staff at the NRC is now committed to these review and licensing actions. To address this additional workload, the NRC has recruited a large number of new Regulatory Staff for dealing with these and other regulatory actions such as the US Fleet of Research and Test Reactors (RTRs). These reactors pose unusual demands on Regulatory Staff since the US Fleet of RTRs, although few (32 Licensed RTRs as of 2010), they represent a broad range of reactor types, operations, and research and training aspects that nuclear reactor power plants (such as the 104 LWRs) do not pose. The US NRC must inspect and regulate all these entities. This paper addresses selected training topics and regulatory activities provided US NRC Inspectors for US RTRs. (author)

Non-aqueous removal of sodium from reactor components

Energy Technology Data Exchange (ETDEWEB)

Welch, F H; Steele, O P [Rockwell International, Atomics International Division, Canoga Park (United States)

1978-08-01

Reactor components from sodium-cooled systems. whether radioactive or not, must have the sodium removed before they can be safely handled for 1) disposal, 2) examination and test, or 3) decontamination, repair, and requalification. In the latter two cases, the sodium must be removed in a manner which will not harm the component. and prevent future use. Two methods for sodium removal using non-aqueous techniques have been studied extensively in the U.S.A. in the past few years: the Alcohol Process, which uses a fully denatured ethanol to react away the sodium; and the Evaporative Process, which uses heat and vacuum to evaporate the sodium from the component.

Non-aqueous removal of sodium from reactor components

International Nuclear Information System (INIS)

Welch, F.H.; Steele, O.P.

1978-01-01

Reactor components from sodium-cooled systems. whether radioactive or not, must have the sodium removed before they can be safely handled for 1) disposal, 2) examination and test, or 3) decontamination, repair, and requalification. In the latter two cases, the sodium must be removed in a manner which will not harm the component. and prevent future use. Two methods for sodium removal using non-aqueous techniques have been studied extensively in the U.S.A. in the past few years: the Alcohol Process, which uses a fully denatured ethanol to react away the sodium; and the Evaporative Process, which uses heat and vacuum to evaporate the sodium from the component

KfK, Institute for Reactor Components. Results of research and development activities in 1989

International Nuclear Information System (INIS)

1990-03-01

R and D activities at IRB (Institut fuer Reaktorbauelemente - Institute for Reactor Components) are dedicated to thermodynamics and fluid dynamics. Emphasis is on the design of nuclear reactor and fusion reactor components. Environmental engineering was added recently. Most activities are applications-oriented. Fundamental investigations focus on energy research and energy technology. The activities are carried out in the framework of different projects (PKF/nuclear fusion, PSF/nuclear safety, PSU/pollution control). Points of main effort are the development of basic liquid-metal-cooled blanket solutions, investigations on natural convection in reactor tanks, and the cooling properties of future containments for pressurized water reactors in the case of nuclear fusion accidents. (orig./GL) [de

Surface modification method for reactor incore structural component

International Nuclear Information System (INIS)

Obata, Minoru; Sudo, Akira.

1996-01-01

A large number of metal or ceramic small spheres accelerated by pressurized air are collided against a surface of a reactor incore structures or a welded surface of the structural components, and then finishing is applied by polishing to form compression stresses on the surface. This can change residual stresses into compressive stress without increasing the strength of the surface. Accordingly, stress corrosion crackings of the incore structural components or welded portions thereof can be prevented thereby enabling to extend the working life of equipments. (T.M.)

An experience of cleaning and decontamination of the BN-350 reactor components

International Nuclear Information System (INIS)

Vasilenko, K.T.; Kochetkov, L.A.; Arkhipov, V.M.; Baklushin, R.P.; Gorlov, A.I.; Kiselev, G.V.; Rezinkin, P.S.; Samarkin, A.A.; Tverdovsky, N.D.

1978-01-01

In the course of start-up, adjustment and operation of the BN-350 reactor there arose a need for cleaning from sodium and decontamination of primary and secondary equipment components. Design schemes of the systems provided for this purpose as well as those specially designed for cleaning of steam generator evaporators are considered. Technological processes of cleaning and decontamination for some reactor components (removable parts of circulating pumps, evaporators, valves) are described, the results are presented. (author)

China: EDF's feedback experience of reactor operating is essential to win international markets

International Nuclear Information System (INIS)

Maillart, H.

2016-01-01

The main assets of EDF on the Chinese nuclear power market is first, its very important feedback experience of reactor operations (EDF cumulates one year of reactor operations every week due to its fleet of 58 reactors), secondly the cooperation with China allowed China to enter nuclear energy in 1983 with the construction of the Daya Bay plant and now to develop its own technology: the CPR-1000 reactor. China is the world leader in terms of nuclear market dynamism with 30 reactors in operation, 24 reactors being built and 40 others planned. A new stage in the Franco-China cooperation would be to share relevant good practices in the managing of both French and Chinese fleets of reactors. EDF has upgraded its commercial international offer, it now proposes to cover all the stages of the nuclear power plant from site selection to plant deconstruction via construction, operation, maintenance and waste management which constitutes a commitment over a 100 year period. (A.C.)

Time to retire : Indicators for aircraft fleets

NARCIS (Netherlands)

Newcamp, Jeffrey; Verhagen, W.J.C.; Curran, R.

2017-01-01

It is well known that aircraft fleets are aging alongside rising operations and support costs. Logisticians and fleet managers who better understand the milestones and timeline of an aging fleet can recognise potential savings. This paper outlines generalised milestones germane to military aircraft

GENERIC, COMPONENT FAILURE DATA BASE FOR LIGHT WATER AND LIQUID SODIUM REACTOR PRAs

Energy Technology Data Exchange (ETDEWEB)

S. A. Eide; S. V. Chmielewski; T. D. Swantz

1990-02-01

A comprehensive generic component failure data base has been developed for light water and liquid sodium reactor probabilistic risk assessments (PRAs) . The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) and the Centralized Reliability Data Organization (CREDO) data bases were used to generate component failure rates . Using this approach, most of the failure rates are based on actual plant data rather than existing estimates .

Three-component homeostasis control

Science.gov (United States)

Xu, Jin; Hong, Hyunsuk; Jo, Junghyo

2014-03-01

Two reciprocal components seem to be sufficient to maintain a control variable constant. However, pancreatic islets adapt three components to control glucose homeostasis. They are α (secreting glucagon), β (insulin), and δ (somatostatin) cells. Glucagon and insulin are the reciprocal hormones for increasing and decreasing blood glucose levels, while the role of somatostatin is unknown. However, it has been known how each hormone affects other cell types. Based on the pulsatile hormone secretion and the cellular interactions, this system can be described as coupled oscillators. In particular, we used the Landau-Stuart model to consider both amplitudes and phases of hormone oscillations. We found that the presence of the third component, δ cell, was effective to resist under glucose perturbations, and to quickly return to the normal glucose level once perturbed. Our analysis suggested that three components are necessary for advanced homeostasis control.

Use of the local-global concept in detecting component vibration in reactors

International Nuclear Information System (INIS)

Al-Ammar, M.A.

1981-01-01

The local-global concept, based on the detector adjoint function, was used to develop the response of a detector to an absorber vibrating in one dimension. A one-dimensional two-group diffusion code was developed to calculate the frequency dependent detector response as a function of detector and absorber positions for the coupled-core UTR-10 reactor. Results from this code indicated the best possible detector and absorber locations, where more detailed calculations were made using a two-group, three-dimensional diffusion code with finite detector and absorber volumes. An experiment was then designed, for the chosen positions, using a vibrating cadmium absorber with a detector on each side. The assembly was placed in the vertical central stringer of the reactor. Investigations were carried out for vibrations in two flux gradients and experimental data were analyzed in the frequency domain using a microcomputer-based data acquisition system. The experimental investigation showed the validity of the local-global concept. A normalized outputs cross power spectral density was developed that correctly predicted the different flux tilts in the two flux gradients. It was also shown that the frequency response of the local component had a wide plateau region. Monitoring the behavior of the normalized cross power spectral density was thought to be a promising indicator for the detection and localization of malfunctioning vibrating components. It might also be used to detect flux irregularities in the vicinity of a vibrating component

Lubrication of nuclear reactor components

International Nuclear Information System (INIS)

Wild, E.; Mack, K.J.

1978-01-01

Safe operation of liquid metal cooled nuclear reactors requires a knowledge of the tribological behaviour of contacting components at high temperatures with slow relative movement at high frictional loads in a chemically aggressive environment. Experiments have been performed on various material combinations in liquid sodium and argon. Because of the small sliding movements, hydrodynamic lubrication is not expected and thus surface finish is an important factor. Tests have been performed on brushed, ground and lapped surfaces. Among the material combinations tested a CrC-coating on a 1.4961 stainless steel substrate performed well. Friction coefficients of 0.35-0.5 in argon and 0.1-1.2 in liquid sodium were recorded. (author)

Manufacturing requirements of reactor assembly components for PFBR (Paper No. 041)

International Nuclear Information System (INIS)

Murty, C.G.K.; Bhoje, S.B.

1987-02-01

This paper enumerates the requirements of 500 MWe Prototype Fast Breeder Reactor (PFBR) components and considering the present state of art of Indian industry an analysis is made on the challenges to be faced in manufacture highlighting the areas needing development. The large sizes and weights of the components coupled with the limitations on shop facilities and ODC transport, demand part of the fabrication to be done at shop and balance assembly work as well as certain assembly machining operations to be done at site work shop. The stringent geometrical tolerances coupled with extensive destructive and non-destructive examinations call for balanced and low heat input welding techniques and special inspection equipment like electronic co-ordinate determination system. The present paper deals with the specific manufacturing problems of the main reactor components. (author)

Development and verification test of integral reactor major components

Energy Technology Data Exchange (ETDEWEB)

Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

1999-03-01

The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability.

Development and verification test of integral reactor major components

International Nuclear Information System (INIS)

Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

1999-03-01

The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability

Investigation of hydrogen generation in a three reactor chemical looping reforming process

International Nuclear Information System (INIS)

Khan, Mohammed N.; Shamim, Tariq

2016-01-01

Highlights: • Three-reactor based chemical looping reforming system for hydrogen production. • Investigation of operating parameters using a system-level model. • Optimum operating conditions for hydrogen production are identified. • Different operating parameters affect the reactor temperatures differently. - Abstract: Chemical looping reforming (CLR) is a relatively new method to produce hydrogen (H_2) and is also used as an energy conversion method for solid, liquid or gaseous fuels. There are various advantages of this method such as inherent carbon dioxide (CO_2) capture, minimal NOx emissions and the H_2 production. In this process, there is no direct contact between the fuel and oxidizer. This method utilizes oxygen from an oxygen carrier which may be a transition metal. The idea is to split the combustion process into three separate sub-processes by employing three separate reactors: air reactor where the oxygen carrier is oxidized by air, fuel reactor where natural gas is oxidized to produce a stream of CO_2 and H_2O and steam reactor where the steam is reduced to produce H_2. In this study, a thermodynamic model with iron oxides as oxygen carrier has been developed using Aspen Plus by employing conservation of mass and energy for all the components of the CLR system. The developed model was employed to investigate the effect of various operating parameters such as mass flow rates of air, fuel, steam and oxygen carrier and fraction of inert material on H_2 and CO_2 production and key reactor temperatures. The results show that the H_2 production increases with the increase in air, fuel and steam flow rates up to a certain limit and stays constant for higher flow rates. The CO_2 production follows a similar trend. Similarly, the H_2 production also increases with the increase in oxide flow rate and fraction of inert material up to a particular value, but then decrease for higher oxide flow rates and inert fractions. Reactor temperatures were also

Commercial vehicle fleet management and information systems. Technical memorandum 3 : ITS fleet management technology resource guide

Science.gov (United States)

1997-05-01

In todays increasingly competitive economic environment, effective management of commercial vehicle fleets is important for all types of carriers and for the trucking industry as a whole. To meet fleet management needs, carriers increasingly are t...

The great refit of the EDF nuclear fleet and post-Fukushima actions. An ambition to extend to 50 or 60 years the exploitation of reactors

International Nuclear Information System (INIS)

Tonnac, Alain de; Perves, Jean-Pierre

2014-01-01

The great refit of the EDF nuclear fleet aims at extending the service lifetime of nuclear reactors. If the extension to 40 years is well undertaken, an extension to 50 or 60 years could only be approved after a return on experience of the Fukushima accident and possible additional requests made by the ASN in reference to facilities ageing. The authors first recall some elements related to the Fukushima accident and outline the fact that the possibility of such a strong tsunami had not been taken in to account by the operator (as it had been for a neighbouring nuclear power station, Onagawa). They also highlight the differences between Japan and France in terms of organisation of the nuclear industry. They also outline technical differences in the design of power stations. Thus, the first lessons learned from the accident seem to support French options. They discuss the evolutions which however are planned for the French nuclear stations after the Fukushima accident. Then, the authors present the great refit and the issue of the extension of the service lifetime of the EDF nuclear fleet. This project is presented as a major challenge for EDF, and implies a management of expenses. They outline the good level of French energy production with respect to greenhouse gas emissions, and briefly discuss the issue of competitiveness for renewable energies

Experimental Facilities for Performance Evaluation of Fast Reactor Components

International Nuclear Information System (INIS)

Chandramouli, S.; Kumar, V.A. Suresh; Shanmugavel, M.; Vijayakumar, G.; Vinod, V.; Noushad, I.B.; Babu, B.; Kumar, G. Padma; Nashine, B.K.; Rajan, K.K.

2013-01-01

Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

Chooz A: a model for the dismantling of water-cooled reactors

International Nuclear Information System (INIS)

Anon.

2017-01-01

The specificity of Chooz-A, the first French pressurized water reactor (PWR), is that the reactor and its major components (pumps, exchangers and cooling circuits) are installed in 2 caves dug out in a hill slope. Chooz-A was operating from 1967 to 1991, in 1993 the fuel was removed and in 2007 EDF received the authorization to dismantle the reactor. In 2012, EDF completed the dismantling of the cave containing the elements of the cooling circuit, a cornerstone was the removing of the four 14 m high steam generators. The dismantling of the pressure vessel began in march 2017, it is the same tools and the same processes that were used for the dismantling of the pressure vessel of the Zorita plant (Spain) in 2016. The end of the Chooz-A dismantling is expected in 2022. The feedback experience will help to standardize practices for the French fleet of PWRs. (A.C.)

Statistical techniques for the identification of reactor component structural vibrations

International Nuclear Information System (INIS)

Kemeny, L.G.

1975-01-01

The identification, on-line and in near real-time, of the vibration frequencies, modes and amplitudes of selected key reactor structural components and the visual monitoring of these phenomena by nuclear power plant operating staff will serve to further the safety and control philosophy of nuclear systems and lead to design optimisation. The School of Nuclear Engineering has developed a data acquisition system for vibration detection and identification. The system is interfaced with the HIFAR research reactor of the Australian Atomic Energy Commission. The reactor serves to simulate noise and vibrational phenomena which might be pertinent in power reactor situations. The data acquisition system consists of a small computer interfaced with a digital correlator and a Fourier transform unit. An incremental tape recorder is utilised as a backing store and as a means of communication with other computers. A small analogue computer and an analogue statistical analyzer can be used in the pre and post computational analysis of signals which are received from neutron and gamma detectors, thermocouples, accelerometers, hydrophones and strain gauges. Investigations carried out to date include a study of the role of local and global pressure fields due to turbulence in coolant flow and pump impeller induced perturbations on (a) control absorbers, (B) fuel element and (c) coolant external circuit and core tank structure component vibrations. (Auth.)

Results of 1989/90 research and development activities at KfK Institute for Reactor Components

International Nuclear Information System (INIS)

1991-03-01

R and D activities at IRB (Institut fuer Reaktorbauelemente - Institute for Reactor Components) are dedicated to thermodynamics and fluid dynamics. Emphasis is on the design of nuclear reactor and fusion reactor components. Environmental engineering was added recently. Most activities are applications-oriented. Fundamental investigations focus on energy research and energy technology. The activities are carried out in the framework of different projects (PKF/nuclear fusion, PSF/nuclear safety, PSU/pollution control). Points of main effort are the development of basic liquid-metal-cooled blanket solutions, investigations on natural convection in reactor ranks, and the cooling properties of future containments for pressurized water reactors in the case of nuclear fusion accidents. (orig./GL) [de

CleanFleet. Final report: Volume 5, employee attitude assessment

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-12-01

The experiences of couriers, operations managers, vehicle handlers (refuelers), and mechanics who drove and/or worked with alternative fuel vehicles, and the attitudes and perceptions of people with these experiences, are examined. Five alternative fuels studied in the CleanFleet project are considers& compressed natural gas, propane gas, California Phase 2 reformulated gasoline, M-85, and electricity. The three major areas of interest include comparative analysis of issues such as health, safety and vehicle performance, business issues encompassing several facets of station operations, and personal commentary and opinions about the CleanFleet project and the alterative fuels. Results of the employee attitude assessment are presented as both statistical and qualitative analysis.

Remaining useful life estimation in heterogeneous fleets working under variable operating conditions

International Nuclear Information System (INIS)

Al-Dahidi, Sameer; Di Maio, Francesco; Baraldi, Piero; Zio, Enrico

2016-01-01

The availability of condition monitoring data for large fleets of similar equipment motivates the development of data-driven prognostic approaches that capitalize on the information contained in such data to estimate equipment Remaining Useful Life (RUL). A main difficulty is that the fleet of equipment typically experiences different operating conditions, which influence both the condition monitoring data and the degradation processes that physically determine the RUL. We propose an approach for RUL estimation from heterogeneous fleet data based on three phases: firstly, the degradation levels (states) of an homogeneous discrete-time finite-state semi-markov model are identified by resorting to an unsupervised ensemble clustering approach. Then, the parameters of the discrete Weibull distributions describing the transitions among the states and their uncertainties are inferred by resorting to the Maximum Likelihood Estimation (MLE) method and to the Fisher Information Matrix (FIM), respectively. Finally, the inferred degradation model is used to estimate the RUL of fleet equipment by direct Monte Carlo (MC) simulation. The proposed approach is applied to two case studies regarding heterogeneous fleets of aluminium electrolytic capacitors and turbofan engines. Results show the effectiveness of the proposed approach in predicting the RUL and its superiority compared to a fuzzy similarity-based approach of literature. - Highlights: • The prediction of the remaining useful life for heterogeneous fleets is addressed. • A data-driven prognostics approach based on a Markov model is proposed. • The proposed approach is applied to two different heterogeneous fleets. • The results are compared with those obtained by a fuzzy similarity-based approach.

Sodium components cleaning status in the Italian fast reactor program

Energy Technology Data Exchange (ETDEWEB)

De Luca, B [CNEN-RIT/MAT - Laboratorio Sviluppo Processi - C.S.N. Cassacia, Rome (Italy); Labanti, V [CNEN-DRV, Bologna (Italy); Mennucci, M [NIRA, Genoa (Italy)

1978-08-01

As a consequence of the Italian Fast Reactor Development, mainly aimed to the PEC project and to the participation in the French Superphenix project, it is of increasing importance to set up a reliable method for specific reactor components and related test loops. The first problem was the cleaning of the PEC fuelling machine. In order to perform the routine maintenance of the machine an alcohol cleaning method based on the use of 2-butoxyethanol-NN dimethylformamide mixture has been proposed.

Aging management of PWR reactor internals in U.S. plants

International Nuclear Information System (INIS)

Amberge, K.J.; Demma, A.

2015-01-01

This paper describes the development, aging management strategies and inspection results of the Pressurized Water Reactor (PWR) vessel internals inspection and evaluation guidelines. The goal of these guidelines is to provide PWR owners with robust aging management strategies to monitor degradation of internals components to support life extension as well as the current period of operation and power up-rate activities. The implementation of these guidelines began in 2010 within the U.S. PWR fleet and several examinations have been performed since. Examples of inspection results are presented for selected vessel internals components and are compared with simulation results. In summary, to date there have been no observations of austenitic stainless steel stress corrosion cracking (SCC), which is consistent with expectations based on the current understanding of the mechanism. Observations of irradiation assisted stress corrosion cracking (IASCC) have been limited and only found in baffle former bolting. Additionally, no macroscopic effects or global observations of void swelling impacts on general conditions of reactor internal hardware have been observed. (authors)

Temporal Optimization Planning for Fleet Repositioning

DEFF Research Database (Denmark)

Tierney, Kevin; Jensen, Rune Møller

2011-01-01

Fleet repositioning problems pose a high financial bur- den on shipping firms, but have received little attention in the literature, despite their high importance to the shipping industry. Fleet repositioning problems are characterized by chains of interacting activities, but state-of-the-art pla......Fleet repositioning problems pose a high financial bur- den on shipping firms, but have received little attention in the literature, despite their high importance to the shipping industry. Fleet repositioning problems are characterized by chains of interacting activities, but state......-of-the-art planning and scheduling techniques do not offer cost models that are rich enough to represent essential objectives of these problems. To this end, we introduce a novel framework called Temporal Optimization Planning (TOP). TOP uses partial order planning to build optimization models associated...

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

International Nuclear Information System (INIS)

Gregg, R.; Hesketh, K.

2013-01-01

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

Specialists' meeting on heat exchanging components of gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1984-07-01

The objective of the Meeting sponsored by IAEA was to provide a forum for the exchange and discussion of technical information related to heat exchanging and heat conducting components for gas-cooled reactors. The technical part of the meeting covered eight subjects: Heat exchanging components for process heat applications, design and requirements, and research and development programs; Status of the design and construction of intermediate He/He exchangers; Design, construction and performance of steam generators; Metallic materials and design codes; Design and construction of valves and hot gas ducts; Description of component test facilities and test results; Manufacturing of heat exchanging components.

Specialists' meeting on heat exchanging components of gas-cooled reactors

International Nuclear Information System (INIS)

1984-01-01

The objective of the Meeting sponsored by IAEA was to provide a forum for the exchange and discussion of technical information related to heat exchanging and heat conducting components for gas-cooled reactors. The technical part of the meeting covered eight subjects: Heat exchanging components for process heat applications, design and requirements, and research and development programs; Status of the design and construction of intermediate He/He exchangers; Design, construction and performance of steam generators; Metallic materials and design codes; Design and construction of valves and hot gas ducts; Description of component test facilities and test results; Manufacturing of heat exchanging components

Materials Degradation in Light Water Reactors: Life After 60,

International Nuclear Information System (INIS)

Busby, Jeremy T; Nanstad, Randy K; Stoller, Roger E; Feng, Zhili; Naus, Dan J

2008-01-01

susceptibility for most components and may introduce new degradation modes. While all components (except perhaps the reactor vessel) can be replaced, it may not be economically favorable. Therefore, understanding, controlling, and mitigating materials degradation processes are key priorities for reactor operation, power uprate considerations, and life extensions. This document is written to give an overview of some of the materials degradation issues that may be key for extend reactor service life. A detailed description of all the possible forms of degradation is beyond the scope of this short paper and has already been described in other documents (for example, the NUREG/CR-6923). The intent of this document is to present an overview of current materials issues in the existing reactor fleet and a brief analysis of the potential impact of extending life beyond 60 years. Discussion is presented in six distinct areas: (1) Reactor pressure vessel; (2) Reactor core and primary systems; (3) Reactor secondary systems; (4) Weldments; (5) Concrete; and (6) Modeling and simulations. Following each of these areas, some research thrust directions to help identify and mitigate lifetime extension issues are proposed. Note that while piping and cabling are important for extended service, these components are discussed in more depth in a separate paper. Further, the materials degradation issues associated with fuel cladding and fuel assemblies are not discussed in this section as these components are replaced periodically and will not influence the overall lifetime of the reactor

Fabrication of nuclear ship reactor MRX model and study on inspection and maintenance of components

International Nuclear Information System (INIS)

Kasahara, Yoshiyuki; Nakazawa, Toshio; Kusunoki, Tsuyoshi; Takahashi, Hiroki; Yoritsune, Tsutomu.

1997-10-01

The MRX (Marine Reactor X) is an integral type small reactor adopting passive safety systems. As for an integral type reactor, primary system components are installed in the reactor vessel. It is therefor important to establish the appropriate procedure for construction, inspection and maintenance, dismauntling, etc., for all components in the reactor vessel as well as in the reactor containment, because inspection space is limited. To study these subjects, a one-fifth model of the MRX was fabricated and operation capabilities were studied. As a result of studies, the following results are obtained. (1) Manufacturing and installing problems of the reactor pressure vessel, the containment vessel and internal components are basically not abserved. (2) Heat transfer tube structures of the steam generator and the heat exchangers of emergency decay heat removal system and containment water cooler were not seen of any problem for fabrication. However, due consideration is required in the detailed design of supports of heat transfer tubes. (3) Further studies should be needed for designs of flange penetrations and leak countermeasures for pipes instrument cables. (4) Arrangements of equipments in the containment should be taken in consideration in detail because the space is narrow. (5) Further discussion is required for installation methods of instruments and cables. (author)

Optimal fleet conversion policy from a life cycle perspective

International Nuclear Information System (INIS)

Hyung Chul Kim; Ross, M.H.; Keoleian, G.A.

2004-01-01

Vehicles typically deteriorate with accumulating mileage and emit more tailpipe air pollutants per mile. Although incentive programs for scrapping old, high-emitting vehicles have been implemented to reduce urban air pollutants and greenhouse gases, these policies may create additional sales of new vehicles as well. From a life cycle perspective, the emissions from both the additional vehicle production and scrapping need to be addressed when evaluating the benefits of scrapping older vehicles. This study explores an optimal fleet conversion policy based on mid-sized internal combustion engine vehicles in the US, defined as one that minimizes total life cycle emissions from the entire fleet of new and used vehicles. To describe vehicles' lifetime emission profiles as functions of accumulated mileage, a series of life cycle inventories characterizing environmental performance for vehicle production, use, and retirement was developed for each model year between 1981 and 2020. A simulation program is developed to investigate ideal and practical fleet conversion policies separately for three regulated pollutants (CO, NMHC, and NO x ) and for CO 2 . According to the simulation results, accelerated scrapping policies are generally recommended to reduce regulated emissions, but they may increase greenhouse gases. Multi- objective analysis based on economic valuation methods was used to investigate trade-offs among emissions of different pollutants for optimal fleet conversion policies. (author)

Comparison of different scenarios for the deployment of fast reactors in France. Results obtained with COSI

International Nuclear Information System (INIS)

Coquelet-Pascal, Christine; Meyer, Maryan; Girieud, Richard

2011-01-01

In the frame of the French Act for waste management, scenarios studies are carried out with the simulation software COSI to compare different options of evolution of the French reactor fleet, possibilities of plutonium recycling, and options of separation and transmutation of minor actinides. The goal of these studies is to evaluate the sustainability of Sodium cooled Fast Reactors (SFR) deployment from plutonium availability viewpoint, as well as the interest of minor actinides transmutation options in SFR and associated impacts (decay heat and toxicity) on cycle facilities and geological storage. Each option has been evaluated separately in dynamic scenarios taking into account the transition between the current nuclear reactor fleet and a generation IV fleet, with the deployment of SFR in replacement of PWR. In this paper, the results of three types of scenarios, in the continuity of the paper of GLOBAL 2009, are discussed: 1 - plutonium recycling in SFR; 2 - plutonium recycling and minor actinides transmutation in SFR; 3 - plutonium recycling and americium transmutation in SFR. MA transmutation in heterogeneous mode (named 'het.') and in homogeneous mode (named 'hom.') are distinguished. By comparison with the previous paper, new scenarios and extended results are added and a global comparison between transmutation performances is done. Scenarios have been optimized to minimize the impacts on fuel cycle: stabilization of plants capacities over 40 years, limitation of minor actinides storage in the case of heterogeneous transmutation, limitation of MA content in the case of homogeneous transmutation. The impact of MA transmutation on decay heat of fresh and spent fuel are also assessed. The selected scenarios bring also elements on the impact of a SFR deployment delayed from 2040 to 2080, with or without MA transmutation, and the impact of an increase in the total power capacity installed by maximizing the SFR share in a symbiotic fleet PWR-SFR. Whatever the

Design of sodium cooled reactor systems and components for maintainability

International Nuclear Information System (INIS)

Carr, R.W.; Charnock, H.O.; McBride, J.P.

1978-09-01

Special maintenability problems associated with the design and operation of sodium cooled reactor plants are discussed. Some examples of both good and bad design practice are introduced from the design of the FFTF plant and other plants. Subjects include design for drainage, cleaning, decontamination, access, component removal, component disassembly and reassembly, remote tooling, jigs, fixtures, and design for minimizing radiation exposure of maintenance personnel. Check lists are included

A multi-species multi-fleet bioeconomic simulation model for the English Channel artisanal fisheries

DEFF Research Database (Denmark)

Ulrich, Clara; Le Gallic, B.; Dunn, M.R.

2002-01-01

Considering the large number of technical interactions between various fishing activities, the English Channel (ICES divisions VIId and VIIe) fisheries may be regarded as one large and diverse multi-country, multi-gear and multi-species artisanal fishery, although rarely studied as such. A whole...... of the model is to study the long-term consequences of various management alternatives on the economic situation of the English and French fleets fishing in the area and on exploited resources. The model describes this feature through the links between three entities on the one hand (stocks, fleets...... and "metiers", i.e. gear x target species x fishing area), and three modules on the other hand (activity, biological production and economics). The model is described and some simulation results are presented. An example simulating a decrease of one fleet segment effort illustrates these technical interactions...

Aviation Safety Modeling and Simulation (ASMM) Propulsion Fleet Modeling: A Tool for Semi-Automatic Construction of CORBA-based Applications from Legacy Fortran Programs

Science.gov (United States)

Sang, Janche

2003-01-01

Within NASA's Aviation Safety Program, NASA GRC participates in the Modeling and Simulation Project called ASMM. NASA GRC s focus is to characterize the propulsion systems performance from a fleet management and maintenance perspective by modeling and through simulation predict the characteristics of two classes of commercial engines (CFM56 and GE90). In prior years, the High Performance Computing and Communication (HPCC) program funded, NASA Glenn in developing a large scale, detailed simulations for the analysis and design of aircraft engines called the Numerical Propulsion System Simulation (NPSS). Three major aspects of this modeling included the integration of different engine components, coupling of multiple disciplines, and engine component zooming at appropriate level fidelity, require relatively tight coupling of different analysis codes. Most of these codes in aerodynamics and solid mechanics are written in Fortran. Refitting these legacy Fortran codes with distributed objects can increase these codes reusability. Aviation Safety s modeling and simulation use in characterizing fleet management has similar needs. The modeling and simulation of these propulsion systems use existing Fortran and C codes that are instrumental in determining the performance of the fleet. The research centers on building a CORBA-based development environment for programmers to easily wrap and couple legacy Fortran codes. This environment consists of a C++ wrapper library to hide the details of CORBA and an efficient remote variable scheme to facilitate data exchange between the client and the server model. Additionally, a Web Service model should also be constructed for evaluation of this technology s use over the next two- three years.

Performance of materials in the component cooling water systems of pressurized water reactors

International Nuclear Information System (INIS)

Lee, B.S.

1993-01-01

The component cooling water (CCW) system provides cooling water to several important loads throughout the plant under all operating conditions. An aging assessment CCW systems in pressurized water reactors (PWRs) was conducted as part of Nuclear Plant Aging Research Program (NPAR) instituted by the US Nuclear Regulatory Commission. This paper presents some of the results on the performances of materials in respect of their application in CCW Systems. All the CCW system failures reported to the Nuclear Plant Reliability Data System (NPRDS) from January 1988 to June 1990 were reviewed; it is concluded that three of the main contributors to CCW system failures are valves, pumps, and heat exchangers. This study identified the modes and causes of failure for these components; most of the causes for the aging-related failures could be related to the performance of materials. Also, in this paper the materials used for these components are reviewed, and there aging mechanisms under CCW system conditions are discussed

The design of a comprehensive microsimulator of household vehicle fleet composition, utilization, and evolution.

Science.gov (United States)

2012-01-01

The report describes a comprehensive vehicle fleet composition, utilization, and evolution : simulator that can be used to forecast household vehicle ownership and mileage by type of : vehicle over time. The components of the simulator are developed ...

Generic component reliability data for research reactors PSA. Final report of the CRP on data acquisition for research reactor PSA. Working material

International Nuclear Information System (INIS)

1993-01-01

The scope of this document is to provide the final reference generic component reliability database information for a variety of research reactor types. As noted in Section 2.1 and Table 3a, many years of component data are represented in the database so that it is expected that the report should provide representative data valid for a number of years. The database provides component failure rates on a time and/or a demand related basis according to the operational modes of the components. At the current time an update of the database is not planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available from the contributing research reactor facilities themselves. As noted in Section 1.2, the report does not include detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussion regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, tabs

Generic component reliability data for research reactors PSA. Final report of the CRP on data acquisition for research reactor PSA. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-12-31

The scope of this document is to provide the final reference generic component reliability database information for a variety of research reactor types. As noted in Section 2.1 and Table 3a, many years of component data are represented in the database so that it is expected that the report should provide representative data valid for a number of years. The database provides component failure rates on a time and/or a demand related basis according to the operational modes of the components. At the current time an update of the database is not planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available from the contributing research reactor facilities themselves. As noted in Section 1.2, the report does not include detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussion regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, tabs.

Manufacture of heavy reactor components with particular considerations to quality assurance

International Nuclear Information System (INIS)

Kreppel, H.; Clausmeyer, H.

1980-01-01

The use of adequate quality assurance measures is one of the most important prerequisites for the manufacture of reactor components. Nature and extent of the quality assurance system at present adopted in the Federal Republic of Germany are illustrated, using the manufacture of a reactor pressure vessel as an example. The system comprises quality organization, planning of all quality assurance measures, quality surveillance through all stages of manufacture and documentation of quality attained. (orig.)

Manufacture of heavy reactor components with particular consideration to quality assurance

International Nuclear Information System (INIS)

Clausmeyer, H.; Kreppel, H.

1977-01-01

The use of adequate quality assurance measures is one of the most important prerequisites for the manufacture of reactor components. Nature and extent of the quality assurance system at present adopted in the Federal Republic of Germany are illustrated, using the manufacture of a reactor pressure vessel as an example. The system comprises quality organization, planning of all quality assurance measures, quality surveillance through all stages of manufacture and documentation of quality attained. (orig.) [de

Manufacture of heavy reactor components with particular consideration to quality assurance

International Nuclear Information System (INIS)

Kreppel, H.; Clausmeyer, H.

1981-01-01

The use of adequate quality assurance measures is one of the most important prerequisites for the manufacture of reactor components. Nature and extent of the quality assurance system at present adopted in the Federal Republic of Germany are illustrated, using the manufacture of a reactor pressure vessel as an example. The system comprises quality organization, planning of all quality assurance measures, quality surveillance through all stages of manufacture and documentation of quality attained. (orig.)

Fuel recycling and 4. generation reactors

International Nuclear Information System (INIS)

Devezeaux de Lavergne, J.G.; Gauche, F.; Mathonniere, G.

2012-01-01

The 4. generation reactors meet the demand for sustainability of nuclear power through the saving of the natural resources, the minimization of the volume of wastes, a high safety standard and a high reliability. In the framework of the GIF (Generation 4. International Forum) France has decided to study the sodium-cooled fast reactor. Fast reactors have the capacity to recycle plutonium efficiently and to burn actinides. The long history of reprocessing-recycling of spent fuels in France is an asset. A prototype reactor named ASTRID could be entered into operation in 2020. This article presents the research program on the sodium-cooled fast reactor, gives the status of the ASTRID project and present the scenario of the progressive implementation of 4. generation reactors in the French reactor fleet. (A.C.)

The main objectives of lifetime management of reactor unit components

International Nuclear Information System (INIS)

Dragunov, Y.; Kurakov, Y.

1998-01-01

The main objectives of the work concerned with life management of reactor components in Russian Federation are as follows: development of regulations in the field of NPP components ageing and lifetime management; investigations of ageing processes; residual life evaluation taking into account the actual state of NPP systems, real loading conditions and number of load cycles, results of in-service inspections; development and implementation of measures for maintaining/enhancing the NPP safety

Laser-Based Maintenance and Repair Technologies for Reactor Components

International Nuclear Information System (INIS)

Masaki Yoda; Naruhiko Mukai; Makoto Ochiai; Masataka Tamura; Satoshi Okada; Katsuhiko Sato; Motohiko Kimura; Yuji Sano; Noboru Saito; Seishi Shima; Tetsuo Yamamoto

2004-01-01

Toshiba has developed various laser-based maintenance and repair technologies and applied them to existing nuclear power plants. Laser-based technology is considered to be the best tool for remote processing in nuclear power plants, and particularly so for the maintenance and repair of reactor core components. Accessibility could be drastically improved by a simple handling system owing to the absence of reactive force against laser irradiation and the flexible optical fiber. For the preventive maintenance, laser peening (LP) technology was developed and applied to reactor components in operating BWR plants. LP is a novel process to improve residual stress from tensile to compressive on material surface layer by irradiating focused high-power laser pulses in water. We have developed a fiber-delivered LP system as a preventive maintenance measure against stress corrosion cracking (SCC). Laser ultrasonic testing (LUT) has a great potential to be applied to the remote inspection of reactor components. Laser-induced surface acoustic wave (SAW) inspection system was developed using a compact probe with a multi-mode optical fiber and an interferometer. The developed system successfully detected a micro slit of 0.5 mm depth on weld metal and heat-affected zone (HAZ). An artificial SCC was also detected by the system. We are developing a new LP system combined with LUT to treat the inner surface of bottom-mounted instruments (BMI) of PWR plants. Underwater laser seal welding (LSW) technology was also developed to apply surface crack. LSW is expected to isolate the crack tip from corrosive water environment and to stop the propagation of the crack. Rapid heating and cooling of the process minimize the heat effect, which extends the applicability to neutron-irradiated material. This paper describes recent advances in the development and application of such laser-based technologies. (authors)

Data base formation for important components of reactor TRIGA MARK II

Energy Technology Data Exchange (ETDEWEB)

Jordan, R; Mavko, B; Kozuh, M [Inst. Jozef Stefan, Ljubljana (Slovenia)

1992-07-01

The paper represents specific data base formation for reactor TRIGA MARK II in Podgorica. Reactor operation data from year 1985 to 1990 were collected. Two groups of collected data were formed. The first group includes components data and the second group covers data of reactor scrams. Time related and demand related models were used for data evaluation. Parameters were estimated by classical method. Similar data bases are useful everywhere where components unavailabilities may have severe drawback. (author) [Slovenian] V referatu smo prikazali raziskavo, v okviru katere smo za raziskovalni reaktor TRIGA MARK II v Podgorici izoblikovali specificno bazo podatkov. Zbrali smo podatke obratovanja reaktorja od leta 1985 do 1990. Rezultate raziskave dogodkov smo razdelili v dve glavni skupini. V prvo spadajo obratovalni podatki o komponentah, v drugo skupino pa spadajo zagoni oz. zaustavitve reaktorja. Podatke smo ovrednotili z modelom v casovnem prostoru in z modelom na zahtevo. Parametre modelov smo dolocili s klasicno metodo. Opisane baze podatkov so uporabne povsod, kjer so lahko posledice nezanesljivega delovanja sistemov velike. [author].

Generic aging management programs for license renewal of BWR reactor coolant systems components

International Nuclear Information System (INIS)

Shah, V.N.; Liu, Y.Y.

2002-01-01

The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

Generic Aging Management Programs for License Renewal of BWR Reactor Coolant System Components

International Nuclear Information System (INIS)

Shah, V.N.; Liu, Y.Y.

2002-01-01

The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

RB research nuclear reactor, Annual report for 1989, I - III

International Nuclear Information System (INIS)

Stefanovic, D.; Pesic, M.; Hadimahmutovic, N.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

1989-12-01

This report is made of three parts. Part one contains a short description of the reactor, reactor operation, incidents, status of reactor equipment and components (nuclear fuel, heavy water, reactor vessel, heavy water circulation system, electronic, electric and mechanical equipment, auxiliary systems and Vax-8250 computer). It includes dosimetry and radiation protection data, personnel and financial data. Second part of this report in concerned with maintenance of reactor components and instrumentation. Part three includes data about reactor utilization during 1989

Irradiation of electronic components and circuits at the Portuguese Research Reactor: Lessons learned

Energy Technology Data Exchange (ETDEWEB)

Marques, J.G.; Ramos, A.R.; Fernandes, A.C.; Santos, J.P. [Centro de Ciencias e Tecnologias Nucleares, Instituto Superior Tecnico, Universidade de Lisboa, Estrada Nacional 10, 2695-066 Bobadela LRS (Portugal)

2015-07-01

The behavior of electronic components and circuits under radiation is a concern shared by the nuclear industry, the space community and the high-energy physics community. Standard commercial components are used as much as possible instead of radiation hard components, since they are easier to obtain and allow a significant reduction of costs. However, these standard components need to be tested in order to determine their radiation tolerance. The Portuguese Research Reactor (RPI) is a 1 MW pool-type reactor, operating since 1961. The irradiation of electronic components and circuits is one area where a 1 MW reactor can be competitive, since the fast neutron fluences required for testing are in most cases well below 10{sup 16} n/cm{sup 2}. A program was started in 1999 to test electronics components and circuits for the LHC facility at CERN, initially using a dedicated in-pool irradiation device and later a beam line with tailored neutron and gamma filters. Neutron filters are essential to reduce the intensity of the thermal neutron flux, which does not produce significant defects in electronic components but produces unwanted radiation from activation of contacts and packages of integrated circuits and also of the printed circuit boards. In irradiations performed within the line-of-sight of the core of a fission reactor there is simultaneous gamma radiation which complicates testing in some cases. Filters can be used to reduce its importance and separate testing with a pure gamma radiation source can contribute to clarify some irradiation results. Practice has shown the need to introduce several improvements to the procedures and facilities over the years. We will review improvements done in the following areas: - Optimization of neutron and gamma filters; - Dosimetry procedures in mixed neutron / gamma fields; - Determination of hardness parameter and 1 MeV-equivalent neutron fluence; - Temperature measurement and control during irradiation; - Follow-up of reactor

Annual report on the state of RB reactor components and equipment, december 1999

International Nuclear Information System (INIS)

Milosevic, M.

1999-12-01

According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1998 the reactor lattice was changed 13 times, meaning that experiments were done with 13 configurations of the reactor core. Total reactor operation amounted to 84 Wh with 40 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment in 1999, reactor operation nd utilization data, plan for regular annual maintenance and refurbishment of reactor equipment and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

Annual report on the state of RB reactor components and equipment, december 1998

International Nuclear Information System (INIS)

Milosevic, M.

1998-12-01

According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1998 the reactor lattice was changed 7 times, meaning that experiments were done with 7 configurations of the reactor core. Total reactor operation amounted to 177.5 Wh with 40 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment in 1998, reactor operation nd utilization data, plan for regular annual maintenance and refurbishment of reactor equipment and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

Repair welding of fusion reactor components. Final technical report

International Nuclear Information System (INIS)

Chin, B.A.; Wang, C.A.

1997-01-01

The exposure of metallic materials, such as structural components of the first wall and blanket of a fusion reactor, to neutron irradiation will induce changes in both the material composition and microstructure. Along with these changes can come a corresponding deterioration in mechanical properties resulting in premature failure. It is, therefore, essential to expect that the repair and replacement of the degraded components will be necessary. Such repairs may require the joining of irradiated materials through the use of fusion welding processes. The present ITER (International Thermonuclear Experimental Reactor) conceptual design is anticipated to have about 5 km of longitudinal welds and ten thousand pipe butt welds in the blanket structure. A recent study by Buende et al. predict that a failure is most likely to occur in a weld. The study is based on data from other large structures, particularly nuclear reactors. The data used also appear to be consistent with the operating experience of the Fast Flux Test Facility (FFTF). This reactor has a fuel pin area comparable with the area of the ITER first wall and has experienced one unanticipated fuel pin failure after two years of operation. The repair of irradiated structures using fusion welding will be difficult due to the entrapped helium. Due to its extremely low solubility in metals, helium will diffuse and agglomerate to form helium bubbles after being trapped at point defects, dislocations, and grain boundaries. Welding of neutron-irradiated type 304 stainless steels has been reported with varying degree of heat-affected zone cracking (HAZ). The objectives of this study were to determine the threshold helium concentrations required to cause HAZ cracking and to investigate techniques that might be used to eliminate the HAZ cracking in welding of helium-containing materials

Metal plutonium conversion to components of nuclear reactor fuel

International Nuclear Information System (INIS)

Subbotin, V.G.; Panov, A.V.; Mashirev, V.P.

2000-01-01

Capabilities of different technologies for plutonium conversion to the fuel components of nuclear reactors are studied. Advantages and shortcomings of aqueous and nonaqueous methods of plutonium treatment are shown. Proposals to combine and coordinate efforts of world scientific and technological community in solving problems concerning plutonium of energetic and weapon origin treatment were put forward. (authors)

Metal plutonium conversion to components of nuclear reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Subbotin, V.G.; Panov, A.V. [Russian Federal Nuclear Center, ALL-Russian Science and Research, Institute of Technical Physics, Snezhinsk (Russian Federation); Mashirev, V.P. [ALL-Russian Science and Research Institute of Chemical Technology, Moscow (Russian Federation)

2000-07-01

Capabilities of different technologies for plutonium conversion to the fuel components of nuclear reactors are studied. Advantages and shortcomings of aqueous and nonaqueous methods of plutonium treatment are shown. Proposals to combine and coordinate efforts of world scientific and technological community in solving problems concerning plutonium of energetic and weapon origin treatment were put forward. (authors)

JSC Case Study: Fleet Experience with E-85 Fuel

Science.gov (United States)

Hummel, Kirck

2009-01-01

JSC has used E-85 as part of an overall strategy to comply with Presidential Executive Order 13423 and the Energy Policy Act. As a Federal fleet, we are required to reduce our petroleum consumption by 2 percent per year, and increase the use of alternative fuels in our vehicles. With the opening of our onsite dispenser in October 2004, JSC became the second federal fleet in Texas and the fifth NASA center to add E-85 fueling capability. JSC has a relatively small number of GSA Flex Fuel fleet vehicles at the present time (we don't include personal vehicles, or other contractor's non-GSA fleet), and there were no reasonably available retail E-85 fuel stations within a 15-minute drive or within five miles (one way). So we decided to install a small 1000 gallon onsite tank and dispenser. It was difficult to obtain a supplier due to our low monthly fuel consumption, and our fuel supplier contract has changed three times in less than five years. We experiences a couple of fuel contamination and quality control issues. JSC obtained good information on E-85 from the National Ethanol Vehicle Coalition (NEVC). We also spoke with Defense Energy Support Center, (DESC), Lawrence Berkeley Laboratory, and US Army Fort Leonard Wood. E-85 is a liquid fuel that is dispensed into our Flexible Fuel Vehicles identically to regular gasoline, so it was easy for our vehicle drivers to make the transition.

Federal Fleet Report

Data.gov (United States)

General Services Administration — Annual report of Federal agencies' motor vehicle fleet data collected in the Federal Automotive Statistical Tool (FAST), a web-based reporting tool cosponsored by...

Experience of partial dismantling and large component removal of light water reactors

International Nuclear Information System (INIS)

Dubourg, M.

1987-01-01

Not any of the French PWR reactors need to be decommissioned before the next decade or early 2000. However, feasibility studies of decommissioning have been undertaken and several dismantling scenarios have been considered including the dismantling of four PWR units and the on-site entombment of the active components into a reactor building for interim disposal. In addition to theoretical evaluation of radwaste volume and activity, several operations of partial dismantling of active components and decontamination activities have been conducted in view of dismantling for both PWR and BWR units. By analyzing the concept of both 900 and 1300 MWe PWR's, it appears that the design improvements taken into account for reducing occupational dose exposure of maintenance personnel and the development of automated tools for performing maintenance and repairs of major components, contribute to facilitate future dismantling and decommissioning operations

A Global Assessment of Fast Reactors in the Future

International Nuclear Information System (INIS)

Devezeaux de Lavergne, J-G.; Mathonnière, G.

2013-01-01

Conclusions: • Fast reactors are the only way to fully achieve nuclear sustainability. • The SFR market cannot exist if a recycling market is not already present. • SFR has many other advantages that clearly outwheight the disadvantages (this trend is increasing). • Large data uncertainties (on uranium resources, world nuclear fleet deployment) return the little precise period at which economic competitiveness will be reached. Anyway, it is most likely to occur sometime in the second half of the century. • However, the market will start earlier, as it is splitted in two phases: before and after the economic competitiveness (this event is in fact country-dependant): – In the first phase 0-2 reactors will be built every year; – In the second phase up to 10-15 reactors will be built every year. • It is rather probable that there will be no more than two or three different Gen IV technologies in the world, because of the market size

41 CFR 102-34.330 - What is the Federal Fleet Report?

Science.gov (United States)

2010-07-01

... Fleet Report? 102-34.330 Section 102-34.330 Public Contracts and Property Management Federal Property... MANAGEMENT Federal Fleet Report § 102-34.330 What is the Federal Fleet Report? The Federal Fleet Report (FFR..., in evaluating the effectiveness of the operation and management of individual fleets to determine...

A three-dimensional transient calculation for the reactor model RAMONA using the COMMIX-2(V) code

International Nuclear Information System (INIS)

Weinberg, D.; Frey, H.H.; Tschoeke, H.

1993-01-01

The safety graded decay heat removal system of the European Fast Reactor needs a high availability. This system operates entirely under natural convection. To guarantee a proper design, experiments are carried out to verify thermal-hydraulic computer codes able to predict precisely local temperature loadings of the components and the reactor tank in the transition region from nominal operation under forced convection to the decay heat removal operation. - With the COMMIX-2 (V) code three-dimensional transient calculations have been performed to simulate experiments in the 360 deg. reactor model RAMONA, scaled 1:20 to the reality with water as simulant fluid for sodium. The computed average and local temperatures as well as the velocity distributions show a good agreement with the experimental results. Further efforts are necessary to reduce the computation time. (orig.)

Solutions against PWSCC in dissimilar welds cracks of reactor components

International Nuclear Information System (INIS)

Schlader, D.; Michaut, B.; Knapp, M

2005-01-01

This article provides a brief overview of the experience accumulated by Framatome ANP in the development and use of repair and mitigation techniques of the PWSCC in dissimilar welds cracks of reactor components. A brief description of the alternatives available to the industry for the solution of this problem for both PWR and BWR reactor types is also included. These solutions have been implemented many times by Framatome ANP in Europe and the US. The article also describes the way the know-how is shared among the different regions of the company in order to offer customer specific solutions. (Author)

Fleet management tools for the utility industry

Energy Technology Data Exchange (ETDEWEB)

Caywood, J. [Terex Telelect, Watertown, SD (United States)

2002-08-01

In order to enable fleet managers to make the appropriate decisions concerning buying, selling, renting, leasing or repairing vehicles, information about ownership and operating cost and utilization are important. Fleet management tools are now available in the heavy construction industry. The author describes one such fleet management services customized for the utility industry: FleetEdge. This system utilizes current and future technologies to enable the managers to take proactive or predictive approaches to fleet management. An example of a company that has opted for this system was reviewed: Reliant Energy-Houston Metro. The data collection process was explained. A mini-computer processor is fitted on the equipment, and the requested monitoring parameters in turn determine the variety of sensors and alarms to be selected. Location can also be monitored through the use of Global Positioning System (GPS) equipment. The parameters normally include low engine oil pressure, high engine operating temperature, high transmission operating temperature, high hydraulic oil temperature, engine hours. Wireless data transfer is effected via cellular or satellite networks. The data is then collected and checked for errors and format at a hub before going to the computer server through the Internet. Information is transmitted from the accounting system to fleet management service through a secure interface. The data is then configured to meet the client's needs. It results in up to date, understandable information to the client. Equipment utilization, downtime and cost by machine or category is easily accessible and printed if required. Fleet equipment status is provided, as well as hourly cost analysis. Other reports are also available. 2 tabs., 2 figs.

Accounting sodium effect in calculation of strength of nuclear reactor components

International Nuclear Information System (INIS)

Nikitin, V.I.

1981-01-01

Accounting methods of liquid sodium effect on long-term strength and creep of structural materials of nuclear reactors are considered. The decrease of pearlite steel strength at the decarburization expense and the decrease of plasticity of austenitic steels at the expense of carburization are noted. The necessity to account thermal transfer of mass is shown. Values of safety factors are presented, they are recommended for the design of reactor component parts with the thickness not less than 1 mm [ru

Current status and prospects of research reactors

International Nuclear Information System (INIS)

Gabaraev, A.B.; Cherepnin, Yu.S.; Tretyakov, I.T.; Khmelshikov, V.V.; Dollezhal, N.A.

2009-01-01

Full text: The first nuclear research reactors (RR) appeared in the 1940s. Their initial purpose was to provide knowledge of the main processes associated with neutron-induced nuclear reactions. Later, the rang of problems addressed expanded substantially. Besides fundamental research in the properties of matter, such reactors are successfully used for dealing with problems in the fields of materials science, nuclear engineering, medicine, isotope production, education, etc. Over the whole period of RR fleet growth, more than six hundred nuclear research facilities were built in 70 countries of the world. As of the end of 2008, the number of Russian research reactors in service was about 20% of the globally operating RR fleet. This paper discusses the current status of the world's RR fleet and describes the capabilities of the experimental reactor facilities existing in Russia. In the 21st century, research reactors will remain in demand to solve scientific and technological problems for innovative development of society. The emerging renaissance of nuclear power, the expanding RR uses for production of isotopes and other applications, the increase in the number of countries willing to use nuclear technologies in energy production, industry and science - all contribute to a rebirth of interest in research reactors. One of the ways to improve the experimental capabilities lies in radical upgrading of the reactor facilities with qualitative changes in the main neutronic characteristics of the core. The associated design approaches are illustrated with the example of the IBR-2M reactor at the JNRI in Dubna. The imperative need restricting the spread of nuclear threat leads us to give up using highly enriched uranium in most research reactors. Development of RR fuel with reduced enrichment in uranium has been one of the priority objectives of NIKIET for many years. This paper presents the latest results obtained along these lines, as applied to pool-type research

The opinion of the IRSN on the safety and radiation protection of the French electronuclear fleet in 2011

International Nuclear Information System (INIS)

2013-01-01

In its first part, this annual report proposes a global assessment of safety and radiation protection in the exploited electronuclear fleet (trends noticed in 2011 in the field of safety and radiation protection). The second part discusses the various events, incidents and anomalies which occurred in 2011: anomalies in studies for safety demonstration, rate unbalance between safety injection lines in 900 MWe reactors, defects in a penetration at the bottom of a vessel of reactor nr 1 of the Gravelines power station, anomalies concerning pipe supports, incident of the 4 May 2011 on the reactor nr 1 of the Tricastin power station, human and organisational failures in reactor control. The last part comments significant evolutions: EDF approach for a continuous safety improvement, control of reactor ageing effects, high room temperature for safety injection pumps, hybrid cores, boil-over risk at the vicinity of Gravelines

RB Research nuclear reactor, Annual report for 2005

International Nuclear Information System (INIS)

Milosevic, M.; Dasic, N.; Ljubenov, V.; Pesic, M.; Nikolic, D; Jevremovic, M.; Minic, D.

2006-01-01

Report on RB reactor operation during 2005 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation during 2005

Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

International Nuclear Information System (INIS)

Griffiths, M.; Coleman, C.E.; Holt, R.A.; Sagat, S.; Urbanic, V.F.; Chow, C.K.

1993-03-01

X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and β-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the α-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of 'breakaway' growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs

Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

Energy Technology Data Exchange (ETDEWEB)

Griffiths, M; Coleman, C E; Holt, R A; Sagat, S; Urbanic, V F [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.; Chow, C K [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Nuclear Research Establishment

1993-03-01

X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and {beta}-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the {alpha}-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of `breakaway` growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs.

CEP action area studies : City of Yellowknife fleet review

International Nuclear Information System (INIS)

2006-01-01

The City of Yellowknife maintains a fleet of vehicles and equipment used for public service. Yellowknife's Community Energy Plan (CEP) has a goal to reduce Yellowknife's fleet fuel consumption and greenhouse gas (GHG) emissions by 20 per cent within 10 years. This report examined current fuel consumption for the City of Yellowknife's fleet based on available data and examined options to reduce fleet fuel consumption. It also quantified the reduction in fuel consumption, GHG emissions, as well as the financial impacts of various options. An overview of the City of Yellowknife's fleet of vehicles and equipment was first provided, as well as calculated annual values for fuel consumption, GHG emissions, and energy consumption for the City of Yellowknife's fleet. Two fleet management case studies were presented. Fuel-efficient vehicles on the market that could provide an opportunity to replace some of the City of Yellowknife's fleet of light vehicles, and bylaw vehicles with fuel-efficient vehicles were discussed. These vehicles include hybrid vehicles, smart cars, compressed natural gas vehicles, and compressed-air vehicles. Equipment maintenance for tires, preventive, synthetic oils, and driver training programs were discussed. Anti-idling campaigns and technologies were also examined. The report concluded with a discussion of renewable fuels such as biodiesel and ethanol blended gasoline. Fleet fuel consumption, GHG and financial impacts were provided for each chapter heading. It was concluded that if all of the measures identified in the report were implemented, an overall decrease in GHG emissions of approximately 15 per cent would be achieved. 15 refs., 8 tabs., 8 figs

Maintenance of Structures, Systems and Components of the RSG-GAS Reactor as an Implementation of BAPETEN Regulation No. 5 Year 2011

International Nuclear Information System (INIS)

Aep Saepudin Catur; Dede Solehudin Fauzi; Djunaidi

2012-01-01

Maintenance activities for structures, systems and components of the reactor, consider prerequisites for the operation of the non power reactor. This activity is intended to ensure that the structures, systems and components function properly. Implementation of reactor maintenance carried out starting from programs establishment, scheduling, maintenance accomplishment and maintenance report. This paper will describe the implementation of reactor maintenance non power as required by BAPETEN regulation no.5, year 2011. By understanding correctly of this regulation, it is expected that maintenance activity of structures, systems and components of the reactor can be successfully performed. The RSG-GAS reactor has implemented various types of reactor maintenance based on BAPETEN regulation no.5 year 2011 properly. As a result failure of the structures, systems and components of the reactor can be minimized then they can be kept reliable. (author)

Implementing Automotive Telematics for Fleet Insurance

Directory of Open Access Journals (Sweden)

Marika Azzopardi

2013-12-01

Full Text Available The advantages of Usage-Based Insurance for automotive covers over conventional rating methods have been discussed in literature for over four decades. Notwithstanding their adoption in insurance markets has been slow. This paper seeks to establish the viability of introducing fleet Telematics-Based Insurance by investigating the perceptions of insurance operators, tracking service providers and corporate fleet owners. At its core, the study involves a SWOT-analysis to appraise Telematics-Based Insurance against conventional premium rating systems. Twenty five key stakeholders in Malta, a country with an insurance industry that represents others in microcosm, were interviewed to develop our analysis. We assert that local insurers have interests in such insurance schemes as enhanced fleet management and monitoring translate into an improved insurance risk. The findings presented here have implications for all stakeholders as we argue that telematics enhance fleet management, TBI improves risk management for insurers and adoption of this technology is dependent on telematics providers increasing the perceived control by insurers over managing this technology.

A three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition in graphite components of advanced gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Morgan, D.O.; Robinson, A.T.; Allen, D.A.; Picton, D.J.; Thornton, D.A. [TCS, Serco, Rutherford House, Olympus Park, Quedgeley, Gloucester, Gloucestershire GL2 4NF (United Kingdom); Shaw, S.E. [EDF Energy, Barnet Way, Barnwood, Gloucester GL4 3RS (United Kingdom)

2011-07-01

This paper describes the development of a three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition (or nuclear heating) throughout the graphite cores of the UK's Advanced Gas-cooled Reactors. Advances in the development of the Monte Carlo radiation transport code MCBEND have enabled the efficient production of detailed fully three-dimensional models that utilise three-dimensional source distributions obtained from Core Follow data supplied by the reactor physics code PANTHER. The calculational approach can be simplified to reduce both the requisite number of intensive radiation transport calculations, as well as the quantity of data output. These simplifications have been qualified by comparison with explicit calculations and they have been shown not to introduce significant systematic uncertainties. Simple calculational approaches are described that allow users of the data to address the effects on neutron damage and nuclear energy deposition predictions of the feedback resulting from the mutual dependencies of graphite weight loss and nuclear energy deposition. (authors)

Activities in the Czech Republic for reactor pressure components lifetime management

International Nuclear Information System (INIS)

Brumovsky, M.

1995-01-01

Preparation of a system of regulatory guides for life assessment of main pressure components, for inspection qualification and demonstration programmes are outlined. Lifetime management programme for NPP with WWER-440/V-213 reactors is described. Figs and tabs

Anomaly Detection in a Fleet of Systems

Data.gov (United States)

National Aeronautics and Space Administration — A fleet is a group of systems (e.g., cars, aircraft) that are designed and manufactured the same way and are intended to be used the same way. For example, a fleet...

Annual report on the state of RB reactor components and equipment, December 1997

International Nuclear Information System (INIS)

Milosevic, M.

1997-12-01

According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1997 the reactor lattice was not changed due to application of the coupled fast-thermal core HERBE. Total reactor operation amounted to 69.5 Wh with 66 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment, plan for forming new HERBE core, plan for regular annual maintenance of the reactor, and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

Fleet management for autonomous vehicles

OpenAIRE

Bsaybes, Sahar; Quilliot, Alain; Wagler, Annegret K.

2016-01-01

The VIPAFLEET project consists in developing models and algorithms for man- aging a fleet of Individual Public Autonomous Vehicles (VIPA). Hereby, we consider a fleet of cars distributed at specified stations in an industrial area to supply internal transportation, where the cars can be used in different modes of circulation (tram mode, elevator mode, taxi mode). One goal is to develop and implement suitable algorithms for each mode in order to satisfy all the requests under an economic point...

Material and component progress within ARCHER for advanced high temperature reactor

International Nuclear Information System (INIS)

Buckthorpe, D.E.; Davies, M.; Pra, F.; Bonnamy, P.; Fokkens, J.; Heijna, M.; Bout, N. de; Vreeling, A.; Bourlier, F.; Lhachemi, D.; Woayehune, A.; Dubiez-le-Goff, S.; Hahner, P.; Futterer, M.; Berka, J.; Kalivodora, J.; Pouchon, M.A.; Schmitt, R.; Homerin, P.; Marsden, B.; Mummery, P.; Mutch, G.; Ponca, D.; Buhl, P.; Hoffmann, M.; Rondet, F.; Pecherty, A.; Baurand, F.; Alenda, F.; Esch, M.; Kohlz, N.; Reed, J.; Fachinger, J.; Klower, Dr.

2014-01-01

The ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R and D) integrated project started in 2011 as part of the European Commission 7. Framework Programme (FP7) for a period of four years to perform High Temperature Reactor technology R and D in support of reactor demonstration. The project consortium encompasses conventional and Nuclear Industry, Utilities, Technical Support Organizations, Research and Development Organizations and Academia. The activities involved contribute to the Generation IV (GIF) International Forum and collaborate with related projects in the US, China, Japan, and the Republic of Korea in cooperation with IAEA and ISTC. This paper addresses the progress of the work on materials and component technologies within ARCHER over the first two years of the project. (authors)

The IRSN's opinion on safety and radiation protection of the French electronuclear fleet in 2013

International Nuclear Information System (INIS)

2014-01-01

The first part of this report proposes an overview of significant events which occurred in 2013 in the French nuclear power plants: safety re-examination process, evolution of safety significant events, aspects related to the containment of radioactive substances (fuel rod cladding, enclosure tightness test), evolution of organisational arrangements during reactor stoppage. The second and main part proposes a presentation of the currently operated French electronuclear reactor fleet, a global assessment of safety and radiation protection, a report of events, incidents and anomalies (and more notably: the pollution of the air compressed production and distribution systems of a Cruas reactor, the issue of resistance to earthquakes of some hardware, and fuel assembly deformations in Nogent 2), and an overview of significant evolutions (control of containment enclosures of 1300 and 1450 MWe reactors, the taking of aggressions into account for the third decennial inspection of 1300 MWe reactors, thermal fatigue in mixing areas, safety and radiation protection management during reactor stoppage, corrosion of the zircaloy 4 cladding of fuel assemblies)

Interfacing the existing cask fleet with the MRS

International Nuclear Information System (INIS)

Doman, J.W.; Hahn, R.E.

1992-01-01

This paper reports that the Department of Energy (DOE) is considering the possibility of using the existing fleet of casks to achieve spent fuel receipt at the Monitored Retrievable Storage (MRS) facility. The existing cask fleet includes the NLI-1/2, the NAC-LWT, the TN-8 (and TN-8L), the TN-9, and the IF-300 casks. Other casks may be available, but their status is not certain. Use of the existing cask fleet at the MRS places additional design requirements on the system, and specifically affects the cask-to-MRS interface. The decision to use the existing cask fleet also places additional demands on training needs and operator certification, and the configuration management system. Some existing cask designs may not be able to mate with a bottom opening hot cell MRS. Use of the existing cask fleet also greatly increases the number of shipments that must be received, to the point that a facility larger than originally envisioned may be required

Alternative fuels for vehicles fleet demonstration program. Final report, volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-06-01

The Alternative Fuels for Vehicles Fleet Demonstration Program (AFV-FDP) was a multiyear effort to collect technical data for use in determining the costs and benefits of alternative-fuel vehicles (AFVs) in typical applications in New York State. This report, Volume 2, includes 13 appendices to Volume 1 that expand upon issues raised therein. Volume 1 provides: (1) Information about the purpose and scope of the AFV-FDP; (2) A summary of AFV-FDP findings organized on the basis of vehicle type and fuel type; (3) A short review of the status of AFV technology development, including examples of companies in the State that are active in developing AFVs and AFV components; and (4) A brief overview of the status of AFV deployment in the State. Volume 3 provides expanded reporting of AFV-FDP technical details, including the complete texts of the brochure Garage Guidelines for Alternative Fuels and the technical report Fleet Experience Survey Report, plus an extensive glossary of AFV terminology. The appendices cover a wide range of issues including: emissions regulations in New York State; production and health effects of ozone; vehicle emissions and control systems; emissions from heavy-duty engines; reformulated gasoline; greenhouse gases; production and characteristics of alternative fuels; the Energy Policy Act of 1992; the Clean Fuel Fleet Program; garage design guidelines for alternative fuels; surveys of fleet managers using alternative fuels; taxes on conventional and alternative fuels; and zero-emission vehicle technology.

Component Degradation Susceptibilities As The Bases For Modeling Reactor Aging Risk

International Nuclear Information System (INIS)

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

2010-01-01

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.

48 CFR 970.2307-1 - Motor vehicle fleet operations.

Science.gov (United States)

2010-10-01

... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Motor vehicle fleet..., Renewable Energy Technologies, Occupational Safety and Drug-Free Work Place 970.2307-1 Motor vehicle fleet... that the Federal motor vehicle fleet will serve as an example and provide a leadership role in the...

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-06-30

from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying

Thermal-structural response of EBR-II major components under reactor operational transients

International Nuclear Information System (INIS)

Chang, L.K.; Lee, M.J.

1983-01-01

Until recently, the LMFBR safety research has been focused primarily on severe but highly unlikely accident, such as hypothetical-core-disruptive accidents (HCDA's), and not enough attention has been given to accident prevention, which is less severe but more likely sequence. The objective of the EBR-II operational reliability testing (ORT) is to demonstrate that the reactor can be designed and operated to prevent accident. A series of mild duty cycles and overpower transients were designed for accident prevention tests. An assessment of the EBR-II major plant components has been performed to assure structural integrity of the reactor plant for the ORT program. In this paper, the thermal-structural response and structural evaluation of the reactor vessel, the reactor-vessel cover, the intermediate heat exchanger (IHX) and the superheater are presented

The production costs of the French nuclear fleet. Synthesis and conclusions. Technical note

International Nuclear Information System (INIS)

2017-09-01

Whereas the French Energy Multi-year Programming (PPE) notably aims at preserving the purchasing power of consumers and the competitiveness of energy prices, this note aims at reporting an assessment of the production cost of the French present nuclear fleet, i.e. the electric power cost at the output of the production installation. The authors first discuss the choice for the methodology of 'cash costs' for the decision to continue of not the exploitation of existing units. They propose a mean assessment of about 33 euro/MWh, state that the present nuclear-based power production in France is profitable, and consider that there is no 'investment wall' to be faced in the near future. They also state that nuclear costs are hardly supposed to increase because they are little sensitive to uranium prices. They consider that dismantling and waste costs are covered at more than 100 per cent by dedicated assets. A technical note comes with this discussion. It discusses cost calculation methods, the assessment of the production cost of French existing reactors (second generation), and some additional elements regarding some cost components

Fleet view of electrified transportation reveals smaller potential to reduce GHG emissions

International Nuclear Information System (INIS)

Meinrenken, Christoph J.; Lackner, Klaus S.

2015-01-01

Highlights: • Novel framework compares GHG of plugins vs. hybrids for any vehicle type/performance. • Fleet GHG can be compared without forecasting market penetrations of vehicle sizes. • GHG/km for pure electrics must account for limited range using novel, modified Utility Factor. • Applied to the US, this points to smaller GHG reduction at fleet level than traditional fleet analyses. - Abstract: Plugin and hybrid vehicles have been shown to offer possible reductions in greenhouse gas (GHG) emissions, depending on grid-carbon-intensity, range and thus life-cycle battery emissions and vehicle weight, and on trip patterns. We present a framework that enables GHG comparisons (well-to-wheel plus storage manufacturing) for three drivetrains (pure-electric, gasoline-hybrid, and plugin-hybrid), both for individual vehicles and for fleets. The framework captures effects of grid- versus vehicle-based electricity generation, grid transmission and charging losses, and manufacturing and carrying batteries. In contrast to previous work, GHG comparisons can be obtained for heterogeneous fleets of varying vehicle sizes (cars, vans, buses, trucks) and performances, without requiring forecasting of such vehicle specs and their respective market penetrations. Further, we show how a novel adaptation of the Utility Factor concept from plug-in-hybrids to mixed fleets of battery-only and gasoline-hybrids is crucial to quantifying battery-only-vehicles’ impact on fleet-wide GHG. To account for regional variations and possible future technology improvements, we show scenarios over a wide spectrum of grid-carbon-intensities (50–1200 g CO 2 e/kW h at wall), vehicle range (∼5–500 km), battery energy densities, and battery life-cycle GHG. Model uncertainties are quantified via sensitivity tests. Applying the framework to trip patterns of US passenger transportation, we find that owing to the interplay of GHG/km, battery size, all-electric range, and trip patterns, GHG

RB Research nuclear reactor, Annual report for 2006

International Nuclear Information System (INIS)

Milosevic, M.; Ljubenov, V.; Pesic, M.; Jevremovic, M.; Minic, D.

2007-01-01

Report on RB reactor operation during 2006 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains detailed data concerned with measurements performed at the RB reactor and a number of significant results obtained

Three-dimensional harmonic control of a nuclear reactor

International Nuclear Information System (INIS)

Potapenko, P.T.

1989-01-01

Algorithms for neutron flux control based on harmonic three-dimensional core are considered. The essence of the considered approach includes determination of harmonics amplitudes by signals self-powered detectors placed in reactor channels and reconstruction of neutron field distribution over the reactor core volume using the data obtained. Neutron field harmonic control is shown to be reduced to independent measurement and calculation of height harmonics in channels using techniques developed for channel power control

Three-dimensional, three-component wall-PIV

Science.gov (United States)

Berthe, André; Kondermann, Daniel; Christensen, Carolyn; Goubergrits, Leonid; Garbe, Christoph; Affeld, Klaus; Kertzscher, Ulrich

2010-06-01

This paper describes a new time-resolved three-dimensional, three-component (3D-3C) measurement technique called wall-PIV. It was developed to assess near wall flow fields and shear rates near non-planar surfaces. The method is based on light absorption according to Beer-Lambert’s law. The fluid containing a molecular dye and seeded with buoyant particles is illuminated by a monochromatic, diffuse light. Due to the dye, the depth of view is limited to the near wall layer. The three-dimensional particle positions can be reconstructed by the intensities of the particle’s projection on an image sensor. The flow estimation is performed by a new algorithm, based on learned particle trajectories. Possible sources of measurement errors related to the wall-PIV technique are analyzed. The accuracy analysis was based on single particle experiments and a three-dimensional artificial data set simulating a rotating sphere.

Periodic Operation of Three-Phase Catalytic Reactors

Czech Academy of Sciences Publication Activity Database

Silveston, P.T.; Hanika, Jiří

2005-01-01

Roč. 82, č. 6 (2005), s. 1105-1142 ISSN 0008-4034 Institutional research plan: CEZ:AV0Z4072921 Keywords : three-phase reactors * trickle bed * periodic operation Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 0.574, year: 2005

RB Research nuclear reactor, Annual report for 2007

International Nuclear Information System (INIS)

Milosevic, M.; Ljubenov, V.; Pesic, M.; Jevremovic, M.; Minic, D.; Sipka, Dj.

2008-01-01

Report on RB reactor operation during 2007 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation during 2007. Majority of measurement were related to spent fuel from the RA reactor, safety of transportation containers and verification of relevant computer codes

Municipal green fleet management in Ontario: best practices manual 2008

Energy Technology Data Exchange (ETDEWEB)

Felder, M.

2008-07-01

When households, institutional and commercial buildings, industry, and especially auto transportation use fossil fuel based energy, they generate carbon dioxide emissions. Considering the environmental concerns and the soaring fuel prices, this could be the opportunity for the municipal fleet operator to assume a leadership position regarding environmental issues and look for new cost efficiencies. To begin, a meticulous examination of the fuel expenditures, a good management approach of fleet operations and a clearly defined system or guidance for lowering fleet fuel costs would be helpful for municipal fleet managers. This document is involved in a pilot program aimed at helping Ontario municipalities understand and promote fleet efficiencies and obtain related environmental benefits. Existing and cost-effective automotive fleet management policies and practices that can mitigate pollution causing global warming are given in this guide. These policies and practices also allow money savings and contribute to a better workplace health and livability of the community. This document is based on the experience of fleet management experts and local governments in Ontario. 54 refs.

A method for evaluation the activity of the reactor components

International Nuclear Information System (INIS)

Gugiu, E.D.; Roth, Cs.

2003-01-01

The ability to predict the radioactivity levels of the reactor components is an important aspect from waste management point of view, as well as from radioprotection purposes. A special case is represented by the research reactors where, one of the major contributions to the radioactivity inventory is due to the experimental devices involved in various research works during reactor life. Generally, aluminum and aluminum alloys are used in manufacturing these devices; as a result, the work presented in this paper is focused on the qualitative and quantitative analysis of the radioactive isotopes contained in these materials. A device used for silicon doping by neutron transmutation that was placed near TRIGA reactor core is investigated. The isotopic content of various samplings drawn from various points of the device was analyzed by gamma spectrometry using a HPGe detector. Computations, using the MCNP5 code, are also performed in order to evaluate the reaction rates for all the isotopes and their reactions. The Monte Carlo simulations are performed for a detailed geometry and material composition of the reactor core and the device. The Origen-S code is also used in order to evaluate the isotopic inventory and the activity values. A detailed analysis regarding the possibility to estimate by computations and/or by gamma spectrometry the activity values of the isotopes which are of interest for decommissioning is presented in the paper. (authors)

Chemical immobilization of fission products reactive with nuclear reactor components

International Nuclear Information System (INIS)

Grossman, L.N.; Kaznoff, A.I.; Clukey, H.V.

1975-01-01

This invention teaches a method of immobilizing deleterious fission products produced in nuclear fuel materials during nuclear fission chain reactions through the use of additives. The additives are disposed with the nuclear fuel materials in controlled quantities to form new compositions preventing attack of reactor components, especially nuclear fuel cld, by the deleterious fission products. (Patent Office Record)

A study on the establishment of component/equipment performance criteria considering Heavy Water Reactor characteristics

Energy Technology Data Exchange (ETDEWEB)

Chang, Keun Sun; Kwon, Young Chul; Lee, Min Kyu; Lee, Yun Soo [Sunmoon Univ., Asan (Korea, Republic of); Chang, Seong Hoong; Ryo, Chang Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Kim, Soong Pyung; Hwnag, Jung Rye; Chung, Chul Kee [Chosun Univ., Gwangju (Korea, Republic of)

2002-03-15

Foreign and domestic technology trends, regulatory requirements, design and researches for heavy water reactors are analyzed. Safety design guides of Canada industry and regulatory documents and consultative documents of Canada regulatory agency are reviewed. Applicability of MOST guidance 16 Revision 'guidance for technical criteria of nuclear reactor facility' is reviewed. Specific performance criteria are established for components and facilities for heavy water reactor.

Control of activation levels to simplify waste management of fusion reactor ferritic steel components

International Nuclear Information System (INIS)

Wiffen, F.W.; Santoro, R.T.

1983-01-01

Activation characteristics of a material for service in the neutron flux of a fusion reactor first wall fall into three areas: waste management, reactor maintenance and repair, and safety. Of these, the waste management area is the most likely to impact the public acceptance of fusion reactors for power generation. The decay of the activity in steels within tens of years could lead to simplified waste disposal or possibly even to materials recycle. Whether or not these can be achieved will be controlled by (1) selection of alloying elements, (2) control of critical impurity elements, and (3) control of cross contamination from other reactor components. Several criteria can be used to judge the acceptability of potential alloying elements in iron, and to define the limits on content of critical impurity elements. One approach is to select and limit alloying additions on the basis of the activity. If material recycle is a goal, N, Al, Ni, Cu, Nb, and Mo must be excluded. If simplified waste storage by shallow land burial is the goal, regulations limit the concentration of only a few isotopes. For first-wall material that will be exposed to 9 MW-y/m 2 service, allowable initial concentration limits include (in at. ppM) Ni < 20,000; Mo < 3650; N < 3650, Cu < 2400; and Nb < 1.0. The other constituent elements of ferritic steels will not be limited. Possible substitutes for the molybdenum normally used to strengthen the steels include W, Ta, Ti, and V

Suitability of Co as an alloy material for components of the primary circuit of HTR reactors

International Nuclear Information System (INIS)

Iniotakis, N.

1977-02-01

For high temperature reactors it is of interest if Co-alloys could be used for the different components of the primary cooling circuit. It has been investigated in detail to what amount the Co-60 created by neutron activation of Co-59 contained in the material of the components could possibly contribute to the contamination of the primary cooling circuit of the reactor. The result of these investigations is compared with the contamination of the cooling circuit by fission and activation products like Co-137, Cs-134, Ag-11om etc. For pebble bed reactors with an OTTO-type fuel management it could be shown that there is no limitation for the use of cobalt in alloys for materials of the components in the primary cooling circuit. The only boundary condition is that the local Thermal Flux at the position of the components should be less than phisub(th) 7 n/cm 2 . sec. (orig.) [de

Structural integrity and management of aging in internal components of BWR reactors

International Nuclear Information System (INIS)

Arganis J, C.R.

2004-01-01

Presently work the bases to apply structural integrity and the handling of the aging of internal components of the pressure vessel of boiling water reactors of water are revised and is carried out an example of structural integrity in the horizontal welding H4 of the encircling one of the core of a reactor, taking data reported in the literature. It is also revised what is required to carry out the handling program or conduct of the aging (AMP). (Author)

RB research nuclear reactor, Annual report for 1984, I - III

International Nuclear Information System (INIS)

Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Zivkovic, B.; Ilic, I.

1984-01-01

The annual report for 1984 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff and financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; development of experimental methods; utilization of the reactor as a radiation source

Review of leakage-flow-induced vibrations of reactor components

International Nuclear Information System (INIS)

Mulcahy, T.M.

1983-05-01

The primary-coolant flow paths of a reactor system are usually subject to close scrutiny in a design review to identify potential flow-induced vibration sources. However, secondary-flow paths through narrow gaps in component supports, which parallel the primary-flow path, occasionally are the excitation source for significant vibrations even though the secondary-flow rates are orders of magnitude smaller than the primary-flow rate. These so-called leakage flow problems are reviewed here to identify design features and excitation sources that should be avoided. Also, design rules of thumb are formulated that can be employed to guide a design, but quantitative prediction of component response is found to require scale-model testing

RB Research nuclear reactor, Annual report for 2004

International Nuclear Information System (INIS)

Dasic, N.; Pesic, M.; Nikolic, D; Jevremovic, M.; Eskirovic, B.

2005-02-01

Report on RB reactor operation during 2004 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. It contains data about reactor operation during previous 8 years. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation, Annex 1. contains data about heavy water degradation, and Annex 2 is the certificate about the crane bridge in the reactor hall

NSSS Component Control System Design of Integral Reactor

International Nuclear Information System (INIS)

Lee, Joon Koo; Kwon, Ho Je; Jeong, Kwong Il; Park, Heui Youn; Koo, In Soo

2005-01-01

MMIS(Man Machine Interface System) of an integral reactor is composed of a Control Room, Plant Protection System, Control System and Monitoring System which are related with the overall plant operation. MMIS is being developed with a new design concept and digital technology to reduce the Human Factor Error and improve the systems' safety, reliability and availability. And CCS(component control system) is also being developed with a new design concept and digital hardware technology A fully digitalized system and design concept are introduced in the NSSS CCS

Seismic behaviour of gas cooled reactor components

International Nuclear Information System (INIS)

1990-08-01

On invitation of the French Government the Specialists' Meeting on the Seismic Behaviour of Gas-Cooled Reactor Components was held at Gif-sur-Yvette, 14-16 November 1989. This was the second Specialists' Meeting on the general subject of gas-cooled reactor seismic design. There were 27 participants from France, the Federal Republic of Germany, Israel, Japan, Spain, Switzerland, the United Kingdom, the Soviet Union, the United States, the CEC and IAEA took the opportunity to present and discuss a total of 16 papers reflecting the state of the art of gained experiences in the field of their seismic qualification approach, seismic analysis methods and of the capabilities of various facilities used to qualify components and verify analytical methods. Since the first meeting, the sophistication and expanded capabilities of both the seismic analytical methods and the test facilities are apparent. The two main methods for seismic analysis, the impedance method and the finite element method, have been computer-programmed in several countries with the capability of each of the codes dependent on the computer capability. The correlations between calculation and tests are dependent on input assumptions such as boundary conditions, soil parameters and various interactions between the soil, the buildings and the contained equipment. The ability to adjust these parameters and match experimental results with calculations was displayed in several of the papers. The expanded capability of some of the new test facilities was graphically displayed by the description of the SAMSON vibration test facility at Juelich, FRG, capable of dynamically testing specimens weighing up to 25 tonnes, and the TAMARIS facility at the CEA laboratories in Gif-sur-Yvette where the largest table is capable of testing specimens weighing up to 100 tonnes. The proceedings of this meeting contain all 16 presented papers. A separate abstract was prepared for each of these papers. Refs, figs and tabs

Component vibration of VVER-reactors - diagnostics and modelling

International Nuclear Information System (INIS)

Altstadt, E.; Scheffler, M.; Weiss, F.-P.

1995-01-01

Flow induced vibrations of reactor pressure vessel (RPV) internals (control element and core barrel motions) at VVER-440 reactors have led to the development of dedicated methods for on-line monitoring. These methods need a certain developed stage of the faults to be detected. To achieve a real sensitive early detection of mechanical faults of RPV internals, a theoretical vibration model was developed based on finite elements. The model comprises the whole primary circuit including the steam generators (SG). By means of that model all eigenfrequencies up to 30 Hz and the corresponding mode shapes were calculated for the normal vibration behaviour. Moreover the shift of eigenfrequencies and of amplitudes due to the degradation or to the failure of internal clamping and spring elements could be investigated, showing that a recognition of such degradations even inside the RPV is possible by pure excore vibration measurements. A true diagnostic, that is the identification of the failed component, might become possible because different faults influence different and well separated eigenfrequencies. (author)

Creep buckling problems in fast reactor components

International Nuclear Information System (INIS)

Ramesh, R.; Damodaran, S.P.; Chellapandi, P.; Chetal, S.C.; Bhoje, S.B.

1995-01-01

Creep buckling analyses for two important components of 500 M We Prototype Fast Breeder Reactor (PFBR), viz. Intermediate Heat Exchanger (IHX) and Inner Vessel (IV), are reported. The INCA code of CASTEM system is used for the large displacement elasto-plastic-creep analysis of IHX shell. As a first step, INCA is validated for a typical benchmark problem dealing with the creep buckling of a tube under external pressure. Prediction of INCA is also compared with the results obtained using Hoff's theory. For IV, considering the prohibitively high computational cost for the actual analysis, a simplified analysis which involves only large displacement elastoplastic buckling analysis is performed using isochronous stress strain curve approach. From both of these analysis is performed using isochronous stress strain curve approach. From both of these analysis, it has been inferred that creep buckling failure mode is not of great concern in the design of PFBR components. It has also been concluded from the analysis that Creep Cross Over Curve given in RCC-MR is applicable for creep buckling failure mode also. (author). 8 refs., 9 figs., 1 tab

Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

International Nuclear Information System (INIS)

Farmer, Mitchell T.; Bunt, R.; Corradini, M.; Ellison, Paul B.; Francis, M.; Gabor, John D.; Gauntt, R.; Henry, C.; Linthicum, R.; Luangdilok, W.; Lutz, R.; Paik, C.; Plys, M.; Rabiti, Cristian; Rempe, J.; Robb, K.; Wachowiak, R.

2015-01-01

The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy's (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

Energy Technology Data Exchange (ETDEWEB)

Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States); Bunt, R. [Southern Nuclear, Atlanta, GA (United States); Corradini, M. [Univ. of Wisconsin, Madison, WI (United States); Ellison, Paul B. [GE Power and Water, Duluth, GA (United States); Francis, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Gabor, John D. [Erin Engineering, Walnut Creek, CA (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Henry, C. [Fauske and Associates, Burr Ridge, IL (United States); Linthicum, R. [Exelon Corp., Chicago, IL (United States); Luangdilok, W. [Fauske and Associates, Burr Ridge, IL (United States); Lutz, R. [PWR Owners Group (PWROG); Paik, C. [Fauske and Associates, Burr Ridge, IL (United States); Plys, M. [Fauske and Associates, Burr Ridge, IL (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Robb, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Wachowiak, R. [Electric Power Research Inst. (EPRI), Knovville, TN (United States)

2015-01-31

The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

Russian Federation: Passive Safety Components for Lead-Cooled Reactor Facilities

International Nuclear Information System (INIS)

Sarkulov, M.K.

2015-01-01

There is a specific range of engineered features used traditionally in nuclear technology. As a rule, main reactivity control systems use conventional active actuators with solid-body control members and/or liquid systems with active injection of liquid absorber. Other operation principles are normally chosen for additional systems. Currently, the traditional approach to improving the reliability of a reactor facility suggests an increase in the number of safety components and systems which provide for mutual assurance or assist each other. There is a great variety of additional reactivity control members designed for the reactor facility control and shutdown, including hydrodynamic members in the form of rods (acting from the coolant flow); floating-type members (absorbers and displacers); storage-type and liquid members (used in separate channels); bulk members (pebble absorber); gas-based members (with a gas absorber); shape-memory members and others. Hydrodynamic systems were introduced at Beloyarsk NPP Units 1 and 2 and proposed for use in other facility designs, Gases and bulk materials have not been commonly accepted: the former because of the high cost of high-efficiency gaseous absorbers, and the latter because of the complecated monitoring of the bulk material position. It is rather difficult and not always necessary to use the same engineering approaches in new lead-cooled reactor facilities as in traditional ones. Similarly to the development of traditional safety systems, passive safety components (devices) shall be designed according to the essential requirements of the nuclear regulations of the Russian Federation

Major Refurbishment of the University of Florida Training Reactor

International Nuclear Information System (INIS)

Joradn, Kelly; Berglund, Matthew; Shea, Brian

2013-01-01

The research reactor fleet is aging with few replacements being built. At the same time the technology for refurbishment of the older reactors has advanced well beyond that of currently installed equipment. The disparity between new and old technology results in an inability to find simple replacements for the older, highly integrated components. The lack of comprehensive guidance for digital equipment adds to the technical problems of installing individual replacement parts. Up to this point, no U. S. facilities have attempted a complete modernization effort because of the time commitment, financial burden, and licensing required for a total upgrade. The University of Florida Training Reactor is tackling this problem with a replacement of nearly all of the major facility sub-systems, including electrical distribution, reactor controls, nuclear instrumentation, security, building management, and environmental controls. This approach offers increased flexibility over the piece-by-piece replacement method by leveraging modern control systems based on global standards and capable of good data interchange with higher levels of redundancy. The UFTR reviewed numerous technologies to arrive at the final system architecture and this 'clean-slate' installation methodology. It is this concept of total system replacement and strict use of modular, open-standards technology that has allowed for a facility design that will be easy to install, maintain, and build upon over time

Major Refurbishment of the University of Florida Training Reactor

Energy Technology Data Exchange (ETDEWEB)

Joradn, Kelly; Berglund, Matthew; Shea, Brian [Univ., of Florida, Florida (United States)

2013-07-01

The research reactor fleet is aging with few replacements being built. At the same time the technology for refurbishment of the older reactors has advanced well beyond that of currently installed equipment. The disparity between new and old technology results in an inability to find simple replacements for the older, highly integrated components. The lack of comprehensive guidance for digital equipment adds to the technical problems of installing individual replacement parts. Up to this point, no U. S. facilities have attempted a complete modernization effort because of the time commitment, financial burden, and licensing required for a total upgrade. The University of Florida Training Reactor is tackling this problem with a replacement of nearly all of the major facility sub-systems, including electrical distribution, reactor controls, nuclear instrumentation, security, building management, and environmental controls. This approach offers increased flexibility over the piece-by-piece replacement method by leveraging modern control systems based on global standards and capable of good data interchange with higher levels of redundancy. The UFTR reviewed numerous technologies to arrive at the final system architecture and this 'clean-slate' installation methodology. It is this concept of total system replacement and strict use of modular, open-standards technology that has allowed for a facility design that will be easy to install, maintain, and build upon over time.

41 CFR 101-39.104-1 - Consolidations into a fleet management system.

Science.gov (United States)

2010-07-01

... fleet management system. 101-39.104-1 Section 101-39.104-1 Public Contracts and Property Management..., TRANSPORTATION, AND MOTOR VEHICLES 39-INTERAGENCY FLEET MANAGEMENT SYSTEMS 39.1-Establishment, Modification, and Discontinuance of Interagency Fleet Management Systems § 101-39.104-1 Consolidations into a fleet management...

CANDU reactor - supporting the nuclear renaissance

International Nuclear Information System (INIS)

Oberth, R.

2010-01-01

'Full text:' The CANDU reactor has proven to be a strong performer in both the Canada, with 22 units constructed in Ontario, New Brunswick and Quebec, as well as in Argentina, Korea, Romania and China where a further nine units are operating and two in the planning stage. The average lifetime capacity factor of the CANDU reactor fleet is 89%. The last seven CANDU projects in Korea, China, and Romania have been completed on budget and on schedule. CANDU reactors have the highest uranium utilization efficiency measures as electricity output per ton of uranium mined. The CANDU fuel channel design using on-power fuelling and a heavy water moderator enables flexible fueling options - from the current natural uranium option to burning uranium recovered from used LWR reactor fuel and even a thorium-based fuel. AECL and the CANDU reactor are poised to participate in the worldwide construction at least 250 new reactors over the next 20 years. (author)

The role of materials in the analysis of fast breeder reactor components

International Nuclear Information System (INIS)

Aubert, Michel; Petrequin, Pierre.

1982-09-01

The analysis of fast breeder reactor components involves the knowledge of certain properties of the materials used. The latter consist of the following: - a body of data required for calculations, including allowable stresses and fatigue strength, as well as the rules applicable to these data, - a number of qualitative requirements serving to guarantee that the quality of the material fully justifies the use of the previously established elements. This duality of concerns is illustrated by some recent examples which occured during the construction of the Super Phenix reactor [fr

The Capabilities and Limitation of Remote Visual Methods to Detect Service-Induced Cracks in Reactor Components

International Nuclear Information System (INIS)

Cumblidge, Stephen E.; Doctor, Steven R.; Anderson, Michael T.

2006-01-01

Since 1977, the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research has funded a multiyear program at the Pacific Northwest National Laboratory (PNNL) to evaluate the reliability and accuracy of nondestructive evaluation (NDE) techniques employed for inservice inspection (ISI). Recently, the U.S. nuclear industry proposed replacing current volumetric and/or surface examinations of certain components in commercial nuclear power plants, as required by ASME Boiler and Pressure Vessel Code Section XI, with a simpler visual testing (VT) method. The advantages of VT are that these tests generally involve much less radiation exposure and examination times than do volumetric examinations such as ultrasonic testing (UT). However, for industry to justify supplementing volumetric methods with VT, and analysis of pertinent issues is needed to support the reliability of VT in determining the structural integrity of reactor components. As piping and pressure vessel components in a nuclear power station are generally underwater and in high radiation field, they need to be examined by VT from a distance with radiation-hardened video systems. Remote visual testing has been used by nuclear utilities to find cracks in pressure vessel cladding in pressurized water reactors, for shrouds in boiling water reactors, and to investigate leaks in piping and reactor components. These visual tests are performed using a wide variety of procedures and equipment. The techniques for remote visual testing use submersible closed-circuit video cameras to examine reactor components and welds. PNNL has conducted a parametric study that examines the important variables that affect the effectiveness of a remote visual test. Tested variables include lighting techniques, camera resolution, camera movement, and magnification. PNNL has also conducted a laboratory test using a commercial visual testing camera system to experimentally determine the ability of the camera system to

A Framework for Creating Value from Fleet Data at Ecosystem Level

Science.gov (United States)

Kinnunen, Sini-Kaisu; Hanski, Jyri; Marttonen-Arola, Salla; Kärri, Timo

2017-09-01

As companies have recently gotten more interested in utilizing the increasingly gathered data and realizing the potential of data analysis, the ability to upgrade data into value for business has been recognized as an advantage. Companies gain competitive advantage if they are able to benefit from the fleet data that is produced both in and outside the boundaries of the company. Benefits of fleet management are based on the possibility to have access to the massive amounts of asset data that can then be utilized e.g. to gain cost savings and to develop products and services. The ambition of the companies is to create value from fleet data but this requires that different actors in ecosystem are working together for a common goal - to get the most value out of fleet data for the ecosystem. In order that this could be possible, we need a framework to meet the requirements of the fleet life-cycle data utilization. This means that the different actors in the ecosystem need to understand their role in the fleet data refining process in order to promote the value creation from fleet data. The objective of this paper is to develop a framework for knowledge management in order to create value from fleet data in ecosystems. As a result, we present a conceptual framework which helps companies to develop their asset management practices related to the fleet of assets.

A FRAMEWORK FOR CREATING VALUE FROM FLEET DATA AT ECOSYSTEM LEVEL

Directory of Open Access Journals (Sweden)

Sini-Kaisu KINNUNEN

2017-07-01

Full Text Available As companies have recently gotten more interested in utilizing the increasingly gathered data and realizing the potential of data analysis, the ability to upgrade data into value for business has been recognized as an advantage. Companies gain competitive advantage if they are able to benefit from the fleet data that is produced both in and outside the boundaries of the company. Benefits of fleet management are based on the possibility to have access to the massive amounts of asset data that can then be utilized e.g. to gain cost savings and to develop products and services. The ambi-tion of the companies is to create value from fleet data but this requires that different actors in ecosystem are working together for a common goal – to get the most value out of fleet data for the ecosystem. In order that this could be possi-ble, we need a framework to meet the requirements of the fleet life-cycle data utilization. This means that the different actors in the ecosystem need to understand their role in the fleet data refining process in order to promote the value creation from fleet data. The objective of this paper is to develop a framework for knowledge management in order to create value from fleet data in ecosystems. As a result, we present a conceptual framework which helps companies to develop their asset management practices related to the fleet of assets.

Mapping sub-surface geostrophic currents from altimetry and a fleet of gliders

Science.gov (United States)

Alvarez, A.; Chiggiato, J.; Schroeder, K.

2013-04-01

Integrating the observations gathered by different platforms into a unique physical picture of the environment is a fundamental aspect of networked ocean observing systems. These are constituted by a spatially distributed set of sensors and platforms that simultaneously monitor a given ocean region. Remote sensing from satellites is an integral part of present ocean observing systems. Due to their autonomy, mobility and controllability, underwater gliders are envisioned to play a significant role in the development of networked ocean observatories. Exploiting synergism between remote sensing and underwater gliders is expected to result on a better characterization of the marine environment than using these observational sources individually. This study investigates a methodology to estimate the three dimensional distribution of geostrophic currents resulting from merging satellite altimetry and in situ samples gathered by a fleet of Slocum gliders. Specifically, the approach computes the volumetric or three dimensional distribution of absolute dynamic height (ADH) that minimizes the total energy of the system while being close to in situ observations and matching the absolute dynamic topography (ADT) observed from satellite at the sea surface. A three dimensional finite element technique is employed to solve the minimization problem. The methodology is validated making use of the dataset collected during the field experiment called Rapid Environmental Picture-2010 (REP-10) carried out by the NATO Undersea Research Center-NURC during August 2010. A marine region off-shore La Spezia (northwest coast of Italy) was sampled by a fleet of three coastal Slocum gliders. Results indicate that the geostrophic current field estimated from gliders and altimetry significantly improves the estimates obtained using only the data gathered by the glider fleet.

Modelling the spatial behaviour of a tropical tuna purse seine fleet.

Directory of Open Access Journals (Sweden)

Tim K Davies

Full Text Available Industrial tuna fisheries operate in the Indian, Atlantic and Pacific Oceans, but concerns over sustainability and environmental impacts of these fisheries have resulted in increased scrutiny of how they are managed. An important but often overlooked factor in the success or failure of tuna fisheries management is the behaviour of fishers and fishing fleets. Uncertainty in how a fishing fleet will respond to management or other influences can be reduced by anticipating fleet behaviour, although to date there has been little research directed at understanding and anticipating the human dimension of tuna fisheries. The aim of this study was to address gaps in knowledge of the behaviour of tuna fleets, using the Indian Ocean tropical tuna purse seine fishery as a case study. We use statistical modelling to examine the factors that influence the spatial behaviour of the purse seine fleet at broad spatiotemporal scales. This analysis reveals very high consistency between years in the use of seasonal fishing grounds by the fleet, as well as a forcing influence of biophysical ocean conditions on the distribution of fishing effort. These findings suggest strong inertia in the spatial behaviour of the fleet, which has important implications for predicting the response of the fleet to natural events or management measures (e.g., spatial closures.

Aging and life extension of major light water reactor components

International Nuclear Information System (INIS)

Shah, V.N.; MacDonald, P.E.

1993-01-01

An understanding of the aging degradation of the major pressurized and boiling water reactor structures and components is given. The design and fabrication of each structure or component is briefly described followed by information on the associated stressors. Interactions between the design, materials and various stressors that cause aging degradation are reviewed. In many cases, aging degradation problems have occurred, and the plant experience to date is analyzed. The discussion summarize the available aging-related information and are supported with extensive references, including references to US Nuclear Regulatory Commission (USNRC) documents, Electric Power Research Institute reports, US and international conference proceedings and other publications

Fishing fleet profiling methodology

National Research Council Canada - National Science Library

Ferraris, Jocelyne

2002-01-01

A fishing fleet profile aims tho assist in understanding the complexity and structure of fisheries from a technical and socio-economic point of view, or from the point of view of fishing strategies...

Human-in-the-Loop simulation in support of long-term sustainability of light water reactors

International Nuclear Information System (INIS)

Hallbert, Bruce P.

2014-01-01

Reliable instrumentation, information, and control systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II and C technology integration. The NPP owners and operators realize that the traditional analog technology represents a significant challenge to sustaining the operation of the current fleet of NPPs. Beyond control systems, new technologies are needed to monitor and characterize the effects of aging and degradation in critical areas of key structures, systems, and components. The objective of the efforts sponsored by the U.S. Department of Energy is to develop, demonstrate, and deploy new digital technologies for II and C architectures and provide monitoring capabilities to ensure the continued safe, reliable, and economic operation of the nation's NPPs. (author)

Hydraulic Hybrid Fleet Vehicle Testing | Transportation Research | NREL

Science.gov (United States)

Hydraulic Hybrid Fleet Vehicle Evaluations Hydraulic Hybrid Fleet Vehicle Evaluations How Hydraulic Hybrid Vehicles Work Hydraulic hybrid systems can capture up to 70% of the kinetic energy that would -pressure reservoir to a high-pressure accumulator. When the vehicle accelerates, fluid in the high-pressure

Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Tan, Lizhen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yang, Ying [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sridharan, K. [Univ. of Wisconsin, Madison, WI (United States)

2015-12-01

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe²⁺) irradiation.

Research of application of new material to light water reactor components

International Nuclear Information System (INIS)

Mihara, Tanetoyo

1992-01-01

Advanced Nuclear Equipment Research Institute (ANERI) has been doing the research to apply the new material including metal, fine ceramics and high polymer which were developed and applied in other industries to components and parts of light water reactor for the purpose of Improvement of reliability of components, improvement of efficiency of periodic inspection, improvement of repair and reduction of radiation exposure of worker. This project started upon the sponsorship of Ministry of International Trade and Industry (MITI) by the schedule of FY1985-FY1993 (9 years) and effective results has been obtained. (author)

RB research nuclear reactor, Annual report for 1983, I - III

International Nuclear Information System (INIS)

Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Zivkovic, B.

1983-01-01

The annual report for 1981 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff; financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; utilization of the reactor as a radiation source. It contains the preliminary safety report for operating the reactor with the internal neutron converter and the plan for criticality experiment with the converter

RB Research nuclear reactor, Annual report for 1996, I-IV

International Nuclear Information System (INIS)

Stefanovic, D.; Milosevic, M.; Pesic, M.; Marinkovic, P.; Ilic, R.; Dasic, N.; Milovanovic, S.; Ljubenov, V.; Petronijevic, M.; Jevremovic, M.

1996-12-01

Report on RB reactor operation during 1996 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a list of publications resulting from experiments done at the RB reactor

RB Research nuclear reactor, Annual report for 1995, I-IV

International Nuclear Information System (INIS)

Stefanovic, D.; Milosevic, M.; Pesic, M.; Marinkovic, P.; Ilic, R.; Dasic, N.; Milovanovic, S.; Ljubenov, V.; Petronijevic, M.; Jevremovic, M.

1995-12-01

Report on RB reactor operation during 1995 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor

Packaging, Transportation, and Disposal Logistics for Large Radioactively Contaminated Reactor Decommissioning Components

International Nuclear Information System (INIS)

Lewis, Mark S.

2008-01-01

The packaging, transportation and disposal of large, retired reactor components from operating or decommissioning nuclear plants pose unique challenges from a technical as well as regulatory compliance standpoint. In addition to the routine considerations associated with any radioactive waste disposition activity, such as characterization, ALARA, and manifesting, the technical challenges for large radioactively contaminated components, such as access, segmentation, removal, packaging, rigging, lifting, mode of transportation, conveyance compatibility, and load securing require significant planning and execution. In addition, the current regulatory framework, domestically in Titles 49 and 10 and internationally in TS-R-1, does not lend itself to the transport of these large radioactively contaminated components, such as reactor vessels, steam generators, reactor pressure vessel heads, and pressurizers, without application for a special permit or arrangement. This paper addresses the methods of overcoming the technical and regulatory challenges. The challenges and disposition decisions do differ during decommissioning versus component replacement during an outage at an operating plant. During decommissioning, there is less concern about critical path for restart and more concern about volume reduction and waste minimization. Segmentation on-site is an available option during decommissioning, since labor and equipment will be readily available and decontamination activities are routine. The reactor building removal path is also of less concern and there are more rigging/lifting options available. Radionuclide assessment is necessary for transportation and disposal characterization. Characterization will dictate the packaging methodology, transportation mode, need for intermediate processing, and the disposal location or availability. Characterization will also assist in determining if the large component can be transported in full compliance with the transportation

Surface erosion of fusion reactor components due to radiation blistering and neutron sputtering

International Nuclear Information System (INIS)

Das, S.K.; Kaminsky, M.

1975-01-01

Radiation blistering and neutron sputtering can lead to the surface erosion of fusion reactor components exposed to plasma radiations. Recent studies of methods to reduce the surface erosion caused by these processes are discussed

Nuclear heat source component design considerations for HTGR process heat reactor plant concept

International Nuclear Information System (INIS)

McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

1982-05-01

The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported [via an intermediate heat exchanger (IHX)] to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

SIMODIS - a software package for simulating nuclear reactor components

International Nuclear Information System (INIS)

Guimaraes, Lamartine; Borges, Eduardo M.

2000-01-01

In this paper it is presented the initial development effort in building a nuclear reactor component simulation package. This package was developed to be used in the MATLAB simulation environment. It uses the graphical capabilities from MATLAB and the advantages of compiled languages, as for instance FORTRAN and C ++ . From the MATLAB it takes the facilities for better displaying the calculated results. From the compiled languages it takes processing speed. So far models from reactor core, UTSG and OTSG have been developed. Also, a series a user-friendly graphical interfaces have been developed for the above models. As a by product a set of water and sodium thermal and physical properties have been developed and may be used directly as a function from MATLAB, or by being called from a model, as part of its calculation process. The whole set was named SIMODIS, which stands for SIstema MODular Integrado de Simulacao. (author)

Radiation shielding for fusion reactors

International Nuclear Information System (INIS)

Santoro, R.T.

2000-01-01

Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. (author)

Evolution of the deep-sea fleet that supports Canada's international trade

Science.gov (United States)

2002-01-01

This study identifies the flag-related trends of fleets used in Canada's international sea-borne trade relative to the world fleet during the 15-year period from 1985 to 1999. The goal is to determine if there is any indication that fleets that serve...

Clean Cities Case Study: Barwood Cab Fleet Study Summary

International Nuclear Information System (INIS)

Whalen, P.

1999-01-01

Barwood Cab Fleet Study Summary is the second in a new series called''Alternative Fuel Information Case Studies,'' designed to present real-world experiences with alternative fuels to fleet managers and other industry stakeholders

COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

Directory of Open Access Journals (Sweden)

Phani Kumar Domalapally

2015-04-01

Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

Executive Order 13693: Planning for Federal Sustainability in the Next Decade; Guidance for Federal Agencies on Executive Order 13693 -- Federal Fleet Management

Energy Technology Data Exchange (ETDEWEB)

2017-01-01

The U.S. Department of Energy (DOE) and U.S. General Services Administration (GSA) are issuing comprehensive guidance on the federal fleet requirements of Executive Order (E.O.) 13693, Planning for Federal Sustainability in the Next Decade (E.O. 13693), to help federal agencies subject to the executive order develop an overall approach for reducing total fleet greenhouse gas (GHG) emissions and fleet-wide per-mile GHG emissions, and ensure the approach helps these agencies meet their requirements. Three key GHG emissions reduction strategies - right-sizing fleets to mission, increasing fleet fuel efficiency, and displacing petroleum with alternative fuel use - are essential to meeting the requirements and are discussed further in this document. This guidance document is intended to help agency Chief Sustainability Officers (CSOs) and headquarters fleet managers craft tailored executable plans that achieve the purpose of E.O. 13693. The guidance will assist agencies in completing the first phase of a comprehensive fleet management framework by identifying the strategies each agency will then implement to meet or exceed its requirements.

Radioactive Legacy of the Russian Pacific Fleet Operations. Final Report

Energy Technology Data Exchange (ETDEWEB)

Compton, K. L.; Novikov, V.M.; Parker, F.L.; Sivintsev, Y.U.

2003-03-25

There have been extensive studies of the current and potential environmental impact of Russian Northern fleet activities. However, despite the fact that the total number of ships in both fleets are comparable, there have been very few studies published in the open literature of the impact of the Pacific fleet. This study of the Pacific fleet's impact on neighboring countries was undertaken to partially remedy this lack of analysis. This study is focused on an evaluation of the inventory of major sources of radioactive material associated with the decommissioning of nuclear submarines, and an evaluation of releases to the atmosphere and their long-range (>100km) transboundary transport.

Fleet management performance monitoring.

Science.gov (United States)

2013-05-01

The principle goal of this project was to enhance and expand the analytical modeling methodology previously developed as part of the Fleet Management Criteria: Disposal Points and Utilization Rates project completed in 2010. The enhanced and ex...

Managerial improvement efforts after finding unreported cracks in reactor components

International Nuclear Information System (INIS)

Kawamura, S.

2006-01-01

In 2002 TEPCO found that there were unreported cracks in reactor components, of which inspection records had been falsified. Stress Corrosion Cracking indications found in Core Shrouds and Primary Loop Re-circulation pipes at some plants were removed from the inspection records and not reported to the regulators. Top management of TEPCO took the responsibility and resigned, and recovery was started under the leadership of new management team. First of all, behavioral standards were reconstituted to strongly support safety-first value. Ethics education was introduced and corporate ethics committee was organized with participation of external experts. Independent assessment organization was established to enhance quality assurance. Information became more transparent through Non-conformance Control Program. As for the material management, prevention and mitigation programs for the Stress Corrosion Cracking of reactor components were re-established. In addition to the above immediate recovery actions, long term improvement initiatives have also been launched and driven by our aspiration to excellence in safe operation of nuclear power plants. Vision and core values were set to align the people. Organizational learning was enhanced by benchmark studies, better systematic use of operational experience, self-assessment and external assessment. Based on these foundation blocks and with strong sponsorship from the top management, work processes were analyzed and improved by Peer Groups. (author)

Canisters and nonfuel components at commercial nuclear reactors

International Nuclear Information System (INIS)

Gibbard, K.; Disbrow, J.

1994-01-01

This paper discusses detailed data on canisters and nonfuel components (NFC) at US commercial nuclear power reactors. A wide variety of NFC have been reported on the Form RW-859, open-quotes Nuclear Fuel Dataclose quotes survey. They may have been integral with an assembly, noncanistered in baskets, destined for disposal as low-level radioactive waste, or stored in canisters. Similarly, data on the family of canistered spent nuclear fuel (SNF) in storage pools was compiled. Approximately 85 percent of the 40,194 pieces of nonfuel assembly (NFA) hardware reported were integral with an assembly. This represents data submitted by 95 of the 107 reactors in 10 generic assembly classes. In addition, a total of 286 canisters have been reported as being in storage pools as of December 31, 1992. However, an additional 264 open baskets were also reported to contain miscellaneous SNF and nonfuel materials, garbage and debris. All of these 286 canisters meet the dimensional envelope requirements specified for disposal for open-quotes standard fuelclose quotes under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (10 CFR 961); most of the baskets do not

48 CFR 970.5223-5 - DOE motor vehicle fleet fuel efficiency.

Science.gov (United States)

2010-10-01

... and Contract Clauses for Management and Operating Contracts 970.5223-5 DOE motor vehicle fleet fuel..., insert the following clause in contracts providing for Contractor management of the motor vehicle fleet... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false DOE motor vehicle fleet...

Guidance for Federal Agencies on Executive Order 13693 - Federal Fleet Management

Energy Technology Data Exchange (ETDEWEB)

None, None

2017-01-02

Document contains guidance on the federal fleet requirements of Executive Order 13693: Planning for Federal Sustainability in the Next Decade and helps federal agencies subject to the executive order develop an overall approach for reducing total fleet greenhouse gas (GHG) emissions and fleet-wide per-mile GHG emissions.

Status of ageing for reactor components in HANARO

International Nuclear Information System (INIS)

Cho, Yeong Garp; Wu, Jong Sup; Lee, Jung Hee; Ryu, Jeong Soo; Choung, Yun Hang; Jun, Byung Jin

2005-01-01

This paper summarizes the ageing status, the history of the performance and maintenance, and the ageing management program for the reactor structure, shutoff units and control absorber units of HANARO which has been operated for 10 years. From the ageing point of view, the results of the visual inspections of the core components, a deformation measurement of the core inner shell, and wear measurements of the fuel channels are described. The histories of the maintenance, performance and drop cycles were evaluated for the shutoff units and control absorber units. Also there is a summary for the lifetime extension program for the shutoff and control absorber units

SDIN a new information system for the EDF's fleet of French nuclear reactors

International Nuclear Information System (INIS)

Corre, Y.; Sasseigne, P.; Herodin, J.M.; Leclerq, J.

2013-01-01

The SDIN is the new information system that allows the management of operating and maintenance activities of EDF's fleet of nuclear power plants. This new system relies on 6 main softwares that allow: 1) the management of operations and maintenance, 2) the management of documentation, 3) the management of activities during unit outages, 4) the management of diagrams, schemes and designs, 5) the reporting of activities, and 6) a unified access to the SDIN. SDIN entered into operation in 2012 (6 years after its launching) in a testing phase in the Blayais power plant and in 2 engineering departments. SDIN is expected to improve the plant performance and its standard of safety as well as to prepare the way in terms of adequate technical conditions for operating life extension. (A.C.)

Dynamic loads on reactor vessel components by low pressure waves

International Nuclear Information System (INIS)

Benkert, J.; Mika, C.; Stegemann, D.; Valero, M.

1978-01-01

Starting from the conservation theorems for mass and impulses the code DRUWE has been developed enabling the calculation of dynamic loads of the reactor shell on the basis of simplified assumptions for the first period shortly after rupture. According to the RSK-guidelines it can be assumed that the whole weld size is opened within 15 msec. This time-dependent opening of the fractured plane can be taken into account in the computer program. The calculation is composed in a way that for a reactor shell devided into cross and angle sections the local, chronological pressure and strength curves, the total dynamic load as well as the moments acting on the fastenings of the reactor shell can be calculated. As input data only geometrical details concerning the concept of the pressure vessel and its components as well as the effective subcooling of the fluid are needed. By means of several parameters the program can be operated in a way that the results are available in form of listings or diagrams, respectively, but also as card pile for further examinations, e.g. strength analysis. (orig./RW) [de

Component and system simulation models for High Flux Isotope Reactor

International Nuclear Information System (INIS)

Sozer, A.

1989-08-01

Component models for the High Flux Isotope Reactor (HFIR) have been developed. The models are HFIR core, heat exchangers, pressurizer pumps, circulation pumps, letdown valves, primary head tank, generic transport delay (pipes), system pressure, loop pressure-flow balance, and decay heat. The models were written in FORTRAN and can be run on different computers, including IBM PCs, as they do not use any specific simulation languages such as ACSL or CSMP. 14 refs., 13 figs

Transient three-phase three-component flow. Pt. 3

International Nuclear Information System (INIS)

Kolev, N.I.

1986-05-01

A mathematical model of a transient three-dimensional three-phase three-component flow described by three-velocity fields in porous body is presented. A combination of separated mass and energy equations together with mixture momentum equations for the flow is used. The mixture equations are used in diffusion form with the assumption that the diffusion velocity can be calculated from empirical correlations. An analytical coupling between the governing equations is developed for calculation of the pressure field. The system is discretized semiimplicitly in 3D-cylindrical space and different solution methods for the algebraic problem are presented. Finally, numerical examples and comparisons with experimental data demonstrate that the method presented is a powerful tool for numerical multiphase flow simulation. (orig.) [de

Fleet equipment performance measurement preventive maintenance model : final report.

Science.gov (United States)

2014-04-01

The concept of preventive maintenance is very important in the effective management and deployment of : vehicle fleets. The Texas Department of Transportation (TxDOT) operates a large fleet of on-road and offroad : equipment. Newer engines and vehicl...

Simulation and calculation of three-reactor system of catalytic reforming

International Nuclear Information System (INIS)

Rikalovska, Tatjana; Markovska, Liljana; Meshko, Vera; Poposka, Filimena

1999-01-01

The process of catalytic reforming has been operated for quite a long time, one can not always find real data for the kinetics and thermodynamics of the reactions that take place during the catalytic reforming process in order to facilitate the designing of reactor system or its simulation in a wide:ran e of process parameters. Kinetic and thermodynamic data have been collected for the reactions that take place during the catalytic reforming process. The stress has been pointed on four major reactions: dehydrogenation of naphthenes (aromatization), dehydrocyclization of paraffins and hydrocracking of naphthenes and paraffins. On the base of such a kinetic model, the reforming process has been described with a system of differential equations. For the purpose of solving these equations computer programs for simulation of a three-reactor system for adiabatic operation of the reactors. The computer simulation of the mathematical model of this three-reactor system has been accomplished by use of the ISIM-dynamic simulator. The results obtained out of the simulation agree very good with the data of the real process of catalytic reforming in OKTA Crude Oil Refinery in Skopje, Macedonia. (Author)

41 CFR 101-39.104 - Notice of establishment of a fleet management system.

Science.gov (United States)

2010-07-01

... of a fleet management system. 101-39.104 Section 101-39.104 Public Contracts and Property Management..., TRANSPORTATION, AND MOTOR VEHICLES 39-INTERAGENCY FLEET MANAGEMENT SYSTEMS 39.1-Establishment, Modification, and Discontinuance of Interagency Fleet Management Systems § 101-39.104 Notice of establishment of a fleet management...

Excellence in fleet combat replacement squadrons: predicting carrier qualification success

OpenAIRE

Smith, Martin P.

1988-01-01

This thesis presents a two-part analysis of excellence criteria for fleet combat replacement squadrons. Part one focuses on the qualitative issues and management techniques identified in outstanding fleet combat replacement squadrons. Part two develops and presents a regression model for predicting a fleet replacement squadron pilot's carrier qualification grade. The model was derived using standard linear regression techniques and the SPSSx software package of the Naval Postgraduate School. ...

Size and transportation capabilities of the existing US cask fleet

International Nuclear Information System (INIS)

Danese, F.L.; Johnson, P.E.; Joy, D.S.

1990-01-01

This study investigates the current spent nuclear fuel cask fleet capability in the United States. In addition, it assesses the degree to which the current fleet would be available, as a contingency, until proposed Office of Civilian Radioactive Waste Management casks become operational. A limited fleet of ten spent fuel transportation casks is found to be readily available for use in Federal waste management efforts over the next decade

Sealing of leaks in the bioshield cooling system of three research reactors

International Nuclear Information System (INIS)

May, R.; Taylor, M.F.

1995-01-01

Water leaks have occurred in the bioshield cooling system of three research reactors. These leaks have been plugged with a sealant based on a blend of a water-based resin and a bentonite-type clay originally developed for sealing similar leaks on power reactors. The mechanism of sealing and development testing of the sealant are described. Application of the sealant to the three reactors sealed the leaks. However, unlike experience with leaks in steel and aluminium systems, some leaks reappeared after several months service - albeit at a leak rate only a very small fraction of the original leak rate. The recurrent defects were readily retreated with sealant and hence, in these instances, the treatment is an effective maintenance procedure for any ageing reactor rather than a permanent cure. (orig.)

The 40 year deadline for the French nuclear stock. Decision process, strengthening option and costs associated with a possible lifetime extension beyond 40 years of EDF reactors

International Nuclear Information System (INIS)

Marignac, Yves

2014-01-01

This report first proposes an overview of the lifetime extension problematic (PLEX) for nuclear plants in a general context (strategy of lifetime extension, and uncertainty and lack of return on experience), and in the case of France (emergence of a strategy of lifetime extension, economic challenges of an extension, uncertainty on investments required by an extension, risk of imposed implicit decisions). It gives a rather detailed overview of the French nuclear reactor fleet, of the status of EDF-operated reactors (construction stages, main lifetime events, regulatory situation of currently exploited reactors) and of their characteristics (common operation principle, main components, main safety functions, main considered severe accidents). It addresses safety issues related to ageing and to the post-Fukushima re-assessment (problematic, robustness increase, limits and weaknesses of additional safety assessments). It presents the currently performed improvements (safety referential, applicable strengthening requirements like safety re-examinations and requirements introduced after Fukushima, applied strengthening prescriptions drawn from additional safety assessments or from the third decennial inspection, with their limitations and perspectives). It proposes strengthening scenarios along with their challenges (in terms of requirement level, procedure and agenda). Three scenarios are proposed (degraded safety, preserved safety, strengthened safety) and analysed in terms of protection against aggressions, diffuse robustness, prevention and management of reactor accident, prevention and management of pool accidents, ultimate control and assistance means. Scenarios are finally compared in terms of costs

Role of subsidies in EU fleet capacity management

DEFF Research Database (Denmark)

Lindebo, Erik

2005-01-01

Fisheries in the European Union (EU) continue to be overexploited by an overcapitalised fishing fleet, despite the best intentions of two decades of capacity adjustment programmes. This paper considers the progress of fishing capacity under the Multi-annual Guidance Programme and examines the imp...... of vessel decommissioning. The Danish fishing fleet case serves as an empirical example in this regard. Comments on the future capacity management regime and the role of subsidies in EU fisheries are offered.......Fisheries in the European Union (EU) continue to be overexploited by an overcapitalised fishing fleet, despite the best intentions of two decades of capacity adjustment programmes. This paper considers the progress of fishing capacity under the Multi-annual Guidance Programme and examines...

Future Fleet Project. What Can We Afford

Science.gov (United States)

2016-12-21

available to maintain the readiness of the fleet. The reduction in readiness spending coupled with continuing demands from COCOMs for assets eroded... coupled with continuing demands from COCOMs for assets eroded the number of ships in ready‑to‑deploy status. CBO’s estimate of $20.2 billion per...less-expensive aircraft carriers, and additional SSNs. Our intent in designing this fleet is to use the money saved in buying less-expensive

Breeder reactors

International Nuclear Information System (INIS)

Gollion, H.

1977-01-01

The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components [fr

Nuclear Reactor RA Safety Report, Vol. 14, Safety protection measures

International Nuclear Information System (INIS)

1986-11-01

Nuclear reactor accidents can be caused by three type of errors: failure of reactor components including (1) control and measuring instrumentation, (2) errors in operation procedure, (3) natural disasters. Safety during reactor operation are secured during its design and construction and later during operation. Both construction and administrative procedures are applied to attain safe operation. Technical safety features include fission product barriers, fuel elements cladding, primary reactor components (reactor vessel, primary cooling pipes, heat exchanger in the pump), reactor building. Safety system is the system for safe reactor shutdown and auxiliary safety system. RA reactor operating regulations and instructions are administrative acts applied to avoid possible human error caused accidents [sr

Three-batch reloading scheme for IRIS reactor extended cycles

International Nuclear Information System (INIS)

Jecmenica, R.; Pevec, D.; Grgic, D.

2004-01-01

To fully exploit the IRIS reactor optimized maintenance, and at the same time improve fuel utilization, a core design enabling a 4-year operating cycle together with a three-batch reloading scheme is desirable. However, this requires not only the increased allowed burnup but also use of fuel with uranium oxide enriched beyond 5%. This paper considers three-batch reloading scheme for a 4-year operating cycle with the assumptions of increased discharge burnup and fuel enrichment beyond 5%. Calculational model of IRIS reactor core has been developed based on FER FA2D code for group constants generation and NRC's PARCS nodal code for global core analysis. Studies have been performed resulting in a preliminary design of a three-batch core configuration for the first cycle. It must be emphasized that this study is outside the current IRIS licensing efforts, which rely on the present fuel technology (enrichment below 5%), but it is of long-term interest for potential future IRIS design upgrades. (author)

Heuristic hybrid game approach for fleet condition-based maintenance planning

International Nuclear Information System (INIS)

Feng, Qiang; Bi, Xiong; Zhao, Xiujie; Chen, Yiran; Sun, Bo

2017-01-01

The condition-based maintenance (CBM) method is commonly used to select appropriate maintenance opportunities according to equipment status over a period of time. The CBM of aircraft fleets is a fleet maintenance planning problem. In this problem, mission requirements, resource constraints, and aircraft statuses are considered to find an optimal strategy set. Given that the maintenance strategies for each aircraft are finite, fleet CBM can be treated as a combinatorial optimization problem. In this study, the process of making a decision on the CBM of military fleets is analyzed. The fleet CBM problem is treated as a two-stage dynamic decision-making problem. Aircraft are divided into dispatch and standby sets; thus, the problem scale is significantly reduced. A heuristic hybrid game (HHG) approach comprising a competition game and a cooperative game is proposed on the basis of heuristic rule. In the dispatch set, a competition game approach is proposed to search for a local optimal strategy matrix. A cooperative game method for the two sets is also proposed to ensure global optimization. Finally, a case study regarding a fleet comprising 20 aircraft is conducted, with the results proving that the approach efficiently generates outcomes that meet the mission risk-oriented schedule requirement. - Highlights: • A new heuristic hybrid game method for fleet condition-based maintenance is proposed. • The problem is simplified by hierarchical solving based on dispatch and standby set. • The local optimal solution is got by competition game algorithm for dispatch set. • The global optimal solution is got by cooperative game algorithm between two sets.

On matrix superpotential and three-component normal modes

Energy Technology Data Exchange (ETDEWEB)

Rodrigues, R. de Lima [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Lima, A.F. de [Universidade Federal de Campina Grande (UFCG), PB (Brazil). Dept. de Fisica; Mello, E.R. Bezerra de; Bezerra, V.B. [Universidade Federal da Paraiba (UFPB), Joao Pessoa, PB (Brazil). Dept. de Fisica]. E-mails: rafael@df.ufcg.edu.br; aerlima@df.ufcg.edu.br; emello@fisica.ufpb.br; valdir@fisica.ufpb.br

2007-07-01

We consider the supersymmetric quantum mechanics(SUSY QM) with three-component normal modes for the Bogomol'nyi-Prasad-Sommerfield (BPS) states. An explicit form of the SUSY QM matrix superpotential is presented and the corresponding three-component bosonic zero-mode eigenfunction is investigated. (author)

A road safety performance indicator for vehicle fleet compatibility.

NARCIS (Netherlands)

Christoph, M. Vis, M.A. Rackliff, L. & Stipdonk, H.

2013-01-01

This paper discusses the development and the application of a safety performance indicator which measures the intrinsic safety of a country's vehicle fleet related to fleet composition. The indicator takes into account both the ‘relative severity’ of individual collisions between different vehicle

The Fleet Numerical Meteorology and Oceanography Center (FNMOC) - Naval

Science.gov (United States)

Meteorology Oceanography Ice You are here: Home › FNMOC FNMOC Logo FNMOC Navigation Meteorology Products Oceanography Products Tropical Applications Climatology and Archived Data Info The Fleet Numerical Meteorology and Oceanography Center (FNMOC) The Fleet Numerical Meteorology and Oceanography Center (FNMOC

Development of guidelines for inelastic analysis in design of fast reactor components

International Nuclear Information System (INIS)

Nakamura, Kyotada; Kasahara, Naoto; Morishita, Masaki; Shibamoto, Hiroshi; Inoue, Kazuhiko; Nakayama, Yasunari

2008-01-01

The interim guidelines for the application of inelastic analysis to design of fast reactor components were developed. These guidelines are referred from 'Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)'. The basic policies of the guidelines are more rational predictions compared with elastic analysis approach and a guarantee of conservative results for design conditions. The guidelines recommend two kinds of constitutive equations to estimate strains conservatively. They also provide the methods for modeling load histories and estimating fatigue and creep damage based on the results of inelastic analysis. The guidelines were applied to typical design examples and their results were summarized as exemplars to support users

A discussion about simplified methodologies for failure assessment of nuclear reactor components

International Nuclear Information System (INIS)

Cruz, J.R.B.; Andrade, A.H.P. de; Landes, J.D.

1996-01-01

Failure of nuclear reactor components like pressure vessels and piping must be avoided for all phases of reactor operation. Especially severe loading conditions come from postulated accident scenarios during which the integrity of the component is required. The use of Fracture Mechanics concepts to investigate the mechanical behavior of flawed structures in the non-linear regime is a complex subject due to the fact that the crack driving force (expressed in terms of J or CTOD) is not /only a function of the cracked geometry, but depends also on the plastic flow properties of the material. Since the numerical solutions by the finite element method are expensive and time consuming, the existence of simplified engineering procedures is of great relevance. These allow a ready identification of the main parameters affecting the crack driving force, and permit a fast and simple evaluation of the structural integrity of the cracked component. This paper presents an overview of the major simplified ductile fracture methodologies that have been proposed in the literature trying to point out their similarities, strong points and negative aspects. Once the best characteristics of each method are identified, they could then be combined to develop a single methodology, one that would be both easy to use and capable of making accurate failure predictions

Fleet-Car Market PENetration Simulator: CPEN user's guide

Energy Technology Data Exchange (ETDEWEB)

Weil, R.

1980-08-01

The purpose of this manual is to assist prospective users in the understanding and execution of Fleet-Car Market PENetration Simulator (CPEN). CPEN is an interactive FORTRAN program whose purpose is to produce estimates of fleet-market-penetration rates of alternative passenger cars that can be described in terms of specific physical and economic attributes. The data were derived from questionnaires distributed to fleet operators affiliated with National Association of Fleet Administrators (NAFA). Besides the NAFA data, CPEN uses 48 variables that are interactively inserted. Complete data-input descriptions are included in the manual along with algorithm and application flowcharts. Examples of complete successful simulator runs are included for alternative program paths. A listing of the computer program and a glossary for CPEN are included.

Proceedings [of the] symposium on zirconium alloys for reactor components

International Nuclear Information System (INIS)

1992-01-01

A two day symposium on zirconium alloys for reactor components (ZARC-91) was organised during 12-13, 1991. There were 6 invited talks and 43 contributed papers in 10 technical sessions. This symposium, took stock of the progress achieved in the development, design, fabrication and quality assurance of zirconium alloy components and emphasized the R and D efforts required for meeting the challenges posed by the rapid growth of nuclear power in our country. Topics like physical metallurgy, corrosion and irradiation behaviour, and in-service inspection were also covered. The proceedings/papers are arranged under the headings: (1)invited talks, (2)fabrication, (3)design requirement, (4)quality assurance, (5)irradiation damage and PIE, (6)corrosion and hydriding, and (7)in-service inspection. (N.B.). refs., figs., tabs

Plug-In Electric Vehicle Handbook for Fleet Managers (Brochure)

Energy Technology Data Exchange (ETDEWEB)

2012-04-01

Plug-in electric vehicles (PEVs) are entering the automobile market and are viable alternatives to conventional vehicles. This guide for fleet managers describes the basics of PEV technology, PEV benefits for fleets, how to select the right PEV, charging a PEV, and PEV maintenance.

Fleet Tools; NREL (National Renewable Energy Laboratory)

Energy Technology Data Exchange (ETDEWEB)

None

2015-04-01

From beverage distributors to shipping companies and federal agencies, industry leaders turn to the National Renewable Energy Laboratory (NREL) to help green their fleet operations. Cost, efficiency, and reliability are top priorities for fleets, and NREL partners know the lab’s portfolio of tools can pinpoint fuel efficiency and emissions-reduction strategies that also support operational the bottom line. NREL is one of the nation’s foremost leaders in medium- and heavy-duty vehicle research and development (R&D) and the go-to source for credible, validated transportation data. NREL developers have drawn on this expertise to create tools grounded in the real-world experiences of commercial and government fleets. Operators can use this comprehensive set of technology- and fuel-neutral tools to explore and analyze equipment and practices, energy-saving strategies, and other operational variables to ensure meaningful performance, financial, and environmental benefits.

Fleet wide motor asset management program

International Nuclear Information System (INIS)

Rosemeier, R.G.; Moxie, J.J.; Mendez, W.E.

2011-01-01

The Institute of Nuclear Power Operations (INPO) performed several studies concerning the effects of motor failures on US power production. Problems with aging and the lack of preventative maintenance were noted as major contributors to this production loss. Westinghouse has developed, and successfully utilized, a computer program to assist plants in the decision making process for maximizing the operational availability of their motor fleet. The program considers motors of a fleet as a single, critical to production asset, and aids the operator in decisions regarding the following: 1.) Purchase of spares. 2.) Rewind and refurbishment prioritization. 3.) Long-term budget forecasting. (author)

Conceptual design of a fission-based integrated test facility for fusion reactor components

International Nuclear Information System (INIS)

Watts, K.D.; Deis, G.A.; Hsu, P.Y.S.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.

1982-01-01

The testing of fusion materials and components in fission reactors will become increasingly important because of lack of fusion engineering test devices in the immediate future and the increasing long-term demand for fusion testing when a fusion reactor test station becomes available. This paper presents the conceptual design of a fission-based Integrated Test Facility (ITF) developed by EG and G Idaho. This facility can accommodate entire first wall/blanket (FW/B) test modules such as those proposed for INTOR and can also accommodate smaller cylindrical modules similar to those designed by Oak Ridge National laboratory (ORNL) and Westinghouse. In addition, the facility can be used to test bulk breeder blanket materials, materials for tritium permeation, and components for performance in a nuclear environment. The ITF provides a cyclic neutron/gamma flux as well as the numerous module and experiment support functions required for truly integrated tests

Electrochemical machining - manufacturing of turbine and reactor components

International Nuclear Information System (INIS)

Otto, K.

1987-01-01

Electrochemical machining is a shaping process for metallic workpieces with complex geometries. Using an electrode corresponding to the negative of the desired shape, the material to be removed is dissolved anodically at erosion rates of up to 10 mm/min. The reproducible shape accuracy lies between 0,02 and 0,08 mm, depending on the machining problem. Surface finishes of less than 18 μm are attained. The hardness of the material has no influence on the metal removal process. The workpiece is not subjected to any thermal stressing during machining. The process is well suited for quantity production of complex parts and is used inter alia for turbine blades and components for nuclear reactors. (orig.) [de

A computer code for Tokamak reactor concepts evaluation

International Nuclear Information System (INIS)

Rosatelli, F.; Raia, G.

1985-01-01

A computer package has been developed which could preliminarily investigate the engineering configuration of a tokamak reactor concept. The code is essentially intended to synthesize, starting from a set of geometrical and plasma physics parameters and the required performances and objectives, three fundamental components of a tokamak reactor core: blanket+shield, TF magnet, PF magnet. An iterative evaluation of the size, power supply and cooling system requirements of these components allows the judgment and the preliminary design optimization on the considered reactor concept. The versatility of the code allows its application both to next generation tokamak devices and power reactor concepts

Evaluation of core compositions for use in breed and burn reactors and limited-separations fuel cycles

International Nuclear Information System (INIS)

Petroski, Robert; Forget, Benoit; Forsberg, Charles

2013-01-01

Highlights: ► Calculated minimum burnup and irradiation damage for B and B reactor compositions. ► Computed doubling time of fuel cycles using B and B reactors and no chemical separations. ► Determined sensitivity of doubling time to using melt refining vs. direct reuse. ► Examined tradeoff between power density and neutronics for different coolants. - Abstract: Previously developed methods for analyzing breed-and-burn (B and B) reactors are applied to a wide range of core compositions. The compositions studied include different fuel types, steel and silicon carbide structure, and sodium, lead/lead bismuth eutectic (LBE), and gas coolants. These compositions are evaluated for use in “minimum burnup” B and B reactors in which it is assumed that blocks comprising the core can be shuffled in all three dimensions to flatten out non-uniformities in burnup. The two figures of merit evaluated are the minimum irradiation damage requirement and reactor fleet doubling time. To minimize irradiation damage, gas coolants perform best, followed by lead/LBE then sodium. High uranium-content metal fuel outperforms compound fuels, and different types of steel are similar and perform slightly better than silicon carbide. Once-through irradiation damage requirements can be surprisingly modest in minimum burnup B and B reactors, with a wide range of compositions viable at irradiation damage levels 50% higher than existing materials data. Doubling times were calculated for a reactor fleet consisting of B and B reactors operating in a limited-separations fuel cycle; i.e., a fuel cycle with no chemical separation of actinides. The effects of different cooling times and removal of fission products using a melt refining process are evaluated. To minimize doubling time, sodium cooled compositions perform best because they are able to achieve core power densities several times larger than compositions using other coolants. A hypothetical sodium-cooled core composition with high

To the application of TV and optical equipment for in-service inspection of reactor vessel and primary circuit component materials

International Nuclear Information System (INIS)

Afonin, Eh.M.; Bachelis, I.M.; Tokarev, E.A.; Yastrebov, V.E.

1985-01-01

Some problems of application of TV and optical equipment for inspection of reactor vessel and primary circuit component materials are considered taking the most widespread WWER-440 type reactor as an example. The most advanrageous objects of the inspection and typical zones of equipment arrangement are shown. Methods and peculiarities of the inspection with the use of TV and optical equipment are presented. Recommendations on rational application of the equipment for the inspection of WWER-440 reactor vessel components are given

Evaluation and comparison of alternative fleet-level selective maintenance models

International Nuclear Information System (INIS)

Schneider, Kellie; Richard Cassady, C.

2015-01-01

Fleet-level selective maintenance refers to the process of identifying the subset of maintenance actions to perform on a fleet of repairable systems when the maintenance resources allocated to the fleet are insufficient for performing all desirable maintenance actions. The original fleet-level selective maintenance model is designed to maximize the probability that all missions in a future set are completed successfully. We extend this model in several ways. First, we consider a cost-based optimization model and show that a special case of this model maximizes the expected value of the number of successful missions in the future set. We also consider the situation in which one or more of the future missions may be canceled. These models and the original fleet-level selective maintenance optimization models are nonlinear. Therefore, we also consider an alternative model in which the objective function can be linearized. We show that the alternative model is a good approximation to the other models. - Highlights: • Investigate nonlinear fleet-level selective maintenance optimization models. • A cost based model is used to maximize the expected number of successful missions. • Another model is allowed to cancel missions if reliability is sufficiently low. • An alternative model has an objective function that can be linearized. • We show that the alternative model is a good approximation to the other models

Dynamic\tmodelling of catalytic three-phase reactors for hydrogenation and oxidation processes

Directory of Open Access Journals (Sweden)

Salmi T.

2000-01-01

Full Text Available The dynamic modelling principles for typical catalytic three-phase reactors, batch autoclaves and fixed (trickle beds were described. The models consist of balance equations for the catalyst particles as well as for the bulk phases of gas and liquid. Rate equations, transport models and mass balances were coupled to generalized heterogeneous models which were solved with respect to time and space with algorithms suitable for stiff differential equations. The aspects of numerical solution strategies were discussed and the procedure was illustrated with three case studies: hydrogenation of aromatics, hydrogenation of aldehydes and oxidation of ferrosulphate. The case studies revealed the importance of mass transfer resistance inside the catalyst pallets as well as the dynamics of the different phases being present in the reactor. Reliable three-phase reactor simulation and scale-up should be based on dynamic heterogeneous models.

Uncertainty in Fleet Renewal: A Case from Maritime Transportation

DEFF Research Database (Denmark)

Pantuso, Giovanni; Fagerholt, Kjetil; Wallace, Stein W.

2016-01-01

This paper addresses the fleet renewal problem and particularly the treatment of uncertainty in the maritime case. A stochastic programming model for the maritime fleet renewal problem is presented. The main contribution is that of assessing whether or not better decisions can be achieved by using...

An Optimization Method for Condition Based Maintenance of Aircraft Fleet Considering Prognostics Uncertainty

Directory of Open Access Journals (Sweden)

Qiang Feng

2014-01-01

Full Text Available An optimization method for condition based maintenance (CBM of aircraft fleet considering prognostics uncertainty is proposed. The CBM and dispatch process of aircraft fleet is analyzed first, and the alternative strategy sets for single aircraft are given. Then, the optimization problem of fleet CBM with lower maintenance cost and dispatch risk is translated to the combinatorial optimization problem of single aircraft strategy. Remain useful life (RUL distribution of the key line replaceable Module (LRM has been transformed into the failure probability of the aircraft and the fleet health status matrix is established. And the calculation method of the costs and risks for mission based on health status matrix and maintenance matrix is given. Further, an optimization method for fleet dispatch and CBM under acceptable risk is proposed based on an improved genetic algorithm. Finally, a fleet of 10 aircrafts is studied to verify the proposed method. The results shows that it could realize optimization and control of the aircraft fleet oriented to mission success.

10 CFR 110.26 - General license for the export of nuclear reactor components.

Science.gov (United States)

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF... Denmark Finland France Germany Greece Indonesia Ireland Italy Japan Latvia Lithuania Luxembourg...

Selection of hardfacing material for components of the Indian Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Bhaduri, A.K.; Indira, R.; Albert, S.K.; Rao, B.P.S.; Jain, S.C.; Asokkumar, S.

2004-01-01

Nickel-base hardfacing alloys have been chosen to replace cobalt-base alloys as hardfacing material for components of the Indian Prototype Fast Breeder Reactor, for minimising the dose rate to personnel during maintenance and decommissioning, and to reduce the shielding thickness required for component handling. Induced activity, dose rate and shielding computations showed that replacing cobalt-base alloys with nickel-base alloys for hardfacing of components would result in a marked reduction in both the dose rate from the components and the thickness of lead handling flasks. Long-term ageing studies on the nickel-base hardface deposits on austenitic stainless steel showed that the hardface deposit would retain adequate hardness at the end of the components' design service-life of 40 years of exposure at 823 K

Reliability Prediction Of System And Component Of Process System Of RSG-GAS Reactor

International Nuclear Information System (INIS)

Sitorus Pane, Jupiter

2001-01-01

The older the reactor the higher the probability of the system and components suffer from loss of function or degradation. This phenomenon occurred because of wear, corrosion, and fatigue. Study on component reliability was generally performed deterministically and statistically. This paper would describe an analysis of using statistical method, i.e. regression Cox, in order to predict the reliability of the components and their environmental influence's factors. The result showed that the dynamics, non safety related, and mechanic components have higher risk of failure, whereas static, safety related, and electric have lower risk of failures. The relative risk value for variable of components dynamics, quality, dummy 1 and dummy 2 are of 1.54, 1.59, 1.50, and 0.83 compare to other components type with each variable. Component with the higher risk have lower reliability than lower one

Component-Level Prognostics Health Management Framework for Passive Components - Advanced Reactor Technology Milestone: M2AT-15PN2301043

Energy Technology Data Exchange (ETDEWEB)

Ramuhalli, Pradeep; Roy, Surajit; Hirt, Evelyn H.; Prowant, Matthew S.; Pitman, Stan G.; Tucker, Joseph C.; Dib, Gerges; Pardini, Allan F.

2015-06-19

This report describes research results to date in support of the integration and demonstration of diagnostics technologies for prototypical advanced reactor passive components (to establish condition indices for monitoring) with model-based prognostics methods. Achieving this objective will necessitate addressing several of the research gaps and technical needs described in previous technical reports in this series.

Development of a three dimension multi-physics code for molten salt fast reactor

International Nuclear Information System (INIS)

Cheng Maosong; Dai Zhimin

2014-01-01

Molten Salt Reactor (MSR) was selected as one of the six innovative nuclear reactors by the Generation IV International Forum (GIF). The circulating-fuel in the can-type molten salt fast reactor makes the neutronics and thermo-hydraulics of the reactor strongly coupled and different from that of traditional solid-fuel reactors. In the present paper: a new coupling model is presented that physically describes the inherent relations between the neutron flux, the delayed neutron precursor, the heat transfer and the turbulent flow. Based on the model, integrating nuclear data processing, CAD modeling, structured and unstructured mesh technology, data analysis and visualization application, a three dimension steady state simulation code system (MSR3DS) for the can-type molten salt fast reactor is developed and validated. In order to demonstrate the ability of the code, the three dimension distributions of the velocity, the neutron flux, the delayed neutron precursor and the temperature were obtained for the simplified MOlten Salt Advanced Reactor Transmuter (MOSART) using this code. The results indicate that the MSR3DS code can provide a feasible description of multi-physical coupling phenomena in can-type molten salt fast reactor. Furthermore, the code can well predict the flow effect of fuel salt and the transport effect of the turbulent diffusion. (authors)

Size and transportation capabilities of the existing U.S. cask fleet

International Nuclear Information System (INIS)

Danese, F.L.; Johnson, P.E.; Joy, D.S.

1990-01-01

This paper investigates the current spent nuclear fuel cask fleet capability in the United States. It assesses the degree to which the current fleet would be available, as a contingency, until proposed Office of Civilian Radioactive Waste Management casks become operational. A limited fleet of ten spent fuel transportation casks is found to be readily available for use in Federal waste management efforts over the next decade

Alternative Fuels Data Center: New Hampshire Fleet Revs up With Natural Gas

Science.gov (United States)

New Hampshire Fleet Revs up With Natural Gas to someone by E-mail Share Alternative Fuels Data Center: New Hampshire Fleet Revs up With Natural Gas on Facebook Tweet about Alternative Fuels Data Center: New Hampshire Fleet Revs up With Natural Gas on Twitter Bookmark Alternative Fuels Data Center

Design and properties of marine reactors and associated R and D

Energy Technology Data Exchange (ETDEWEB)

Gagarinski, A; Ignatiev, V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Devell, L [Studsvik Eco and Safety AB, Nykoeping (Sweden)

1996-05-01

The report is a review of open information available in the USA, UK, France, Russia and other countries on the design and properties of marine reactors and associated R and D. First, a short discussion is given of the milestones and main trends for the development of nuclear-powered ships. Then a brief review is presented of features for ship reactor design. Light water and liquid metal cooled reactor technologies are described and reactor operating experiences for Russian ice-breakers assessed. Traditional and alternative civil uses of submarine and surface shipboard reactor technology in Russia and Japan are also treated. Finally, some problems connected with radioactive waste by the nuclear-powered fleet are briefly considered. 41 refs, 27 figs, 19 tabs.

Design and properties of marine reactors and associated R and D

International Nuclear Information System (INIS)

Gagarinski, A.; Ignatiev, V.; Devell, L.

1996-05-01

The report is a review of open information available in the USA, UK, France, Russia and other countries on the design and properties of marine reactors and associated R and D. First, a short discussion is given of the milestones and main trends for the development of nuclear-powered ships. Then a brief review is presented of features for ship reactor design. Light water and liquid metal cooled reactor technologies are described and reactor operating experiences for Russian ice-breakers assessed. Traditional and alternative civil uses of submarine and surface shipboard reactor technology in Russia and Japan are also treated. Finally, some problems connected with radioactive waste by the nuclear-powered fleet are briefly considered. 41 refs, 27 figs, 19 tabs

Comparison of control systems applied to the handling of radioactive reactor components

International Nuclear Information System (INIS)

Robinson, C.; Harris, E.G.; Dyer, P.C.; Williams, J.G.B.

1985-01-01

The first generation of nuclear power stations have individual reactors each incorporating complete facilities for servicing components and refuelling. In the later designs, each power station has two reactors which are connected by a central block. This central block contains one set of facilities to service both reactors, but to improve the station capability, some of these are to be replicated. The central block incorporates a hoist well which was used during construction for the accessing of complete components. On completion of this work, the physical size of the hoist well is such as to permit the incorporation of additional facilities if these are shown to be operationally and economically desirable. Since a number of years of power operation has elapsed, the advantages of back-fitting to existing fuel-handling facilities has been illustrated. Since the mechanical arrangements and operating procedures are substantially similar for both the original and new handling facilities, the paper will illustrate the control systems provided for each. The configuration of the system is arranged to have two channels of control which complies with the current standard requirements in the United Kingdom. These requirements are more stringent than when the existing facility was designed and constructed, as described in the relevant sections of the paper. The new system has been designed and is being manufactured to comply with the Central Electricity Generating Board standard for nuclear fuel route interlock and control systems. (author)

Initial development of a practical safety audit tool to assess fleet safety management practices.

Science.gov (United States)

Mitchell, Rebecca; Friswell, Rena; Mooren, Lori

2012-07-01

Work-related vehicle crashes are a common cause of occupational injury. Yet, there are few studies that investigate management practices used for light vehicle fleets (i.e. vehicles less than 4.5 tonnes). One of the impediments to obtaining and sharing information on effective fleet safety management is the lack of an evidence-based, standardised measurement tool. This article describes the initial development of an audit tool to assess fleet safety management practices in light vehicle fleets. The audit tool was developed by triangulating information from a review of the literature on fleet safety management practices and from semi-structured interviews with 15 fleet managers and 21 fleet drivers. A preliminary useability assessment was conducted with 5 organisations. The audit tool assesses the management of fleet safety against five core categories: (1) management, systems and processes; (2) monitoring and assessment; (3) employee recruitment, training and education; (4) vehicle technology, selection and maintenance; and (5) vehicle journeys. Each of these core categories has between 1 and 3 sub-categories. Organisations are rated at one of 4 levels on each sub-category. The fleet safety management audit tool is designed to identify the extent to which fleet safety is managed in an organisation against best practice. It is intended that the audit tool be used to conduct audits within an organisation to provide an indicator of progress in managing fleet safety and to consistently benchmark performance against other organisations. Application of the tool by fleet safety researchers is now needed to inform its further development and refinement and to permit psychometric evaluation. Copyright © 2012 Elsevier Ltd. All rights reserved.

New Trends in Robotics for Agriculture: Integration and Assessment of a Real Fleet of Robots

Directory of Open Access Journals (Sweden)

Luis Emmi

2014-01-01

Full Text Available Computer-based sensors and actuators such as global positioning systems, machine vision, and laser-based sensors have progressively been incorporated into mobile robots with the aim of configuring autonomous systems capable of shifting operator activities in agricultural tasks. However, the incorporation of many electronic systems into a robot impairs its reliability and increases its cost. Hardware minimization, as well as software minimization and ease of integration, is essential to obtain feasible robotic systems. A step forward in the application of automatic equipment in agriculture is the use of fleets of robots, in which a number of specialized robots collaborate to accomplish one or several agricultural tasks. This paper strives to develop a system architecture for both individual robots and robots working in fleets to improve reliability, decrease complexity and costs, and permit the integration of software from different developers. Several solutions are studied, from a fully distributed to a whole integrated architecture in which a central computer runs all processes. This work also studies diverse topologies for controlling fleets of robots and advances other prospective topologies. The architecture presented in this paper is being successfully applied in the RHEA fleet, which comprises three ground mobile units based on a commercial tractor chassis.

New Trends in Robotics for Agriculture: Integration and Assessment of a Real Fleet of Robots

Science.gov (United States)

Gonzalez-de-Soto, Mariano; Pajares, Gonzalo

2014-01-01

Computer-based sensors and actuators such as global positioning systems, machine vision, and laser-based sensors have progressively been incorporated into mobile robots with the aim of configuring autonomous systems capable of shifting operator activities in agricultural tasks. However, the incorporation of many electronic systems into a robot impairs its reliability and increases its cost. Hardware minimization, as well as software minimization and ease of integration, is essential to obtain feasible robotic systems. A step forward in the application of automatic equipment in agriculture is the use of fleets of robots, in which a number of specialized robots collaborate to accomplish one or several agricultural tasks. This paper strives to develop a system architecture for both individual robots and robots working in fleets to improve reliability, decrease complexity and costs, and permit the integration of software from different developers. Several solutions are studied, from a fully distributed to a whole integrated architecture in which a central computer runs all processes. This work also studies diverse topologies for controlling fleets of robots and advances other prospective topologies. The architecture presented in this paper is being successfully applied in the RHEA fleet, which comprises three ground mobile units based on a commercial tractor chassis. PMID:25143976

New trends in robotics for agriculture: integration and assessment of a real fleet of robots.

Science.gov (United States)

Emmi, Luis; Gonzalez-de-Soto, Mariano; Pajares, Gonzalo; Gonzalez-de-Santos, Pablo

2014-01-01

Computer-based sensors and actuators such as global positioning systems, machine vision, and laser-based sensors have progressively been incorporated into mobile robots with the aim of configuring autonomous systems capable of shifting operator activities in agricultural tasks. However, the incorporation of many electronic systems into a robot impairs its reliability and increases its cost. Hardware minimization, as well as software minimization and ease of integration, is essential to obtain feasible robotic systems. A step forward in the application of automatic equipment in agriculture is the use of fleets of robots, in which a number of specialized robots collaborate to accomplish one or several agricultural tasks. This paper strives to develop a system architecture for both individual robots and robots working in fleets to improve reliability, decrease complexity and costs, and permit the integration of software from different developers. Several solutions are studied, from a fully distributed to a whole integrated architecture in which a central computer runs all processes. This work also studies diverse topologies for controlling fleets of robots and advances other prospective topologies. The architecture presented in this paper is being successfully applied in the RHEA fleet, which comprises three ground mobile units based on a commercial tractor chassis.

Study on the Safety Classification Criteria of Mechanical Systems and Components for Open Pool-Type Research Reactors

International Nuclear Information System (INIS)

Belal, Al Momani; Jo, Jong Chull

2013-01-01

This paper describes a new compromised safety classification approach based on the comparative study of the different practices in safety classification of mechanical systems and components of open pool-type RRs, which have been adopted by several developed countries in the nuclear power area. It is hoped that the proposed safety classification criteria will be used to develop a harmonized consensus international standard. Different safety classification criteria for systems, structures, and components (SSCs) of nuclear reactors are used among the countries that export or import nuclear reactor technology, which may make the nuclear technology trade and exchange difficult. Thus, such various different approaches of safety classification need to be compromised to establish a global standard. This article proposes practicable optimized criteria for safety classification of SSCs for open pool-type research reactors (RRs)

Study on the Safety Classification Criteria of Mechanical Systems and Components for Open Pool-Type Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Belal, Al Momani [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Jo, Jong Chull [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

This paper describes a new compromised safety classification approach based on the comparative study of the different practices in safety classification of mechanical systems and components of open pool-type RRs, which have been adopted by several developed countries in the nuclear power area. It is hoped that the proposed safety classification criteria will be used to develop a harmonized consensus international standard. Different safety classification criteria for systems, structures, and components (SSCs) of nuclear reactors are used among the countries that export or import nuclear reactor technology, which may make the nuclear technology trade and exchange difficult. Thus, such various different approaches of safety classification need to be compromised to establish a global standard. This article proposes practicable optimized criteria for safety classification of SSCs for open pool-type research reactors (RRs)

Research of three-dimensional transient reactivity feedback in fast reactor

International Nuclear Information System (INIS)

Xu Li; Shi Gong; Ma Dayuan; Yu Hong

2013-01-01

To solve the three-dimensional time-spatial kinetics feedback problems in fast reactor, a mathematical model of the direct reactivity feedback was proposed. Based on the NAS code for fast reactor and the reactivity feedback mechanism, a feedback model which combined the direct reactivity feedback and feedback reflected by the cross section variation was provided for the transient calculation. Furthermore, the fast reactor group collapsing system was added to the code, thus the real time group collapsing calculation could be realized. The isothermal elevated temperature test of CEFR was simulated by using the code. By comparing the calculation result with the test result of the temperature reactivity coefficient, the validity of the model and the code is verified. (authors)

Telematics Framework for Federal Agencies: Lessons from the Marine Corps Fleet

Energy Technology Data Exchange (ETDEWEB)

Hodge, Cabell [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Singer, Mark R. [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

2017-10-24

Executive Order 13693 requires federal agencies to acquire telematics for their light- and medium-duty vehicles as appropriate. This report is intended to help agencies that are deploying telematics systems and seeking to integrate them into their fleet management process. It provides an overview of telematics capabilities, lessons learned from the deployment of telematics in the Marine Corps fleet, and recommendations for federal fleet managers to maximize value from telematics.

Addressing the minimum fleet problem in on-demand urban mobility.

Science.gov (United States)

Vazifeh, M M; Santi, P; Resta, G; Strogatz, S H; Ratti, C

2018-05-01

Information and communication technologies have opened the way to new solutions for urban mobility that provide better ways to match individuals with on-demand vehicles. However, a fundamental unsolved problem is how best to size and operate a fleet of vehicles, given a certain demand for personal mobility. Previous studies 1-5 either do not provide a scalable solution or require changes in human attitudes towards mobility. Here we provide a network-based solution to the following 'minimum fleet problem', given a collection of trips (specified by origin, destination and start time), of how to determine the minimum number of vehicles needed to serve all the trips without incurring any delay to the passengers. By introducing the notion of a 'vehicle-sharing network', we present an optimal computationally efficient solution to the problem, as well as a nearly optimal solution amenable to real-time implementation. We test both solutions on a dataset of 150 million taxi trips taken in the city of New York over one year 6 . The real-time implementation of the method with near-optimal service levels allows a 30 per cent reduction in fleet size compared to current taxi operation. Although constraints on driver availability and the existence of abnormal trip demands may lead to a relatively larger optimal value for the fleet size than that predicted here, the fleet size remains robust for a wide range of variations in historical trip demand. These predicted reductions in fleet size follow directly from a reorganization of taxi dispatching that could be implemented with a simple urban app; they do not assume ride sharing 7-9 , nor require changes to regulations, business models, or human attitudes towards mobility to become effective. Our results could become even more relevant in the years ahead as fleets of networked, self-driving cars become commonplace 10-14 .

Optimising end of generation of Magnox reactors

International Nuclear Information System (INIS)

Hall, D.; Hopper, E.D.A.

2014-01-01

Designing, justifying and gaining regulatory approval for optimised, terminal fuel cycles for the last 4 of the 13 strong Magnox Fleet is described, covering: - constraints set by the plant owner's integrated closure plan, opportunities for innovative fuel cycles while preserving flexibility to respond to business changes; - methods of collectively determining best options for each site; - selected strategies including lower fuel element retention and inter-reactor transfer of fuel; - the required work scope, its technical, safety case and resource challenges and how they were met; - achieving additional electricity generation worth in excess of Pound 1 b from 4 sites (a total of 8 reactors); - the keys to success. (authors)

TQL in the Fleet: From Theory to Practice

Science.gov (United States)

1993-10-01

distribution is unlimited. BEST-AVAILABLE Copy 94-04214 Total Quality Leadership , TQL, Total Quality Kanaginent, 191, i lo of pilo quality management ...703-602-8942). LINDA M. DOHERTY Director Department of the Navy Total Quality Leadership Office TOL In the Fleet: From Theory to Practice...the Fleet: From Theory to Practice ix OVERVIEW ) ver the past year, the Department of the Navy Total Quality Leadership Office (DON TQL Office) has

Advanced Instrumentation, Information, and Control Systems Technologies Research in Support of Light Water Reactors

International Nuclear Information System (INIS)

Hallbert, Bruce P.; Kenneth, Thomas

2014-01-01

The Advanced Instrumentation, Information, and Control (II and C) Systems Technologies Pathway conducts targeted research and development (R and D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals to ensure that legacy analog II and C systems are not life-limiting issues for the LWR fleet, and to implement digital II and C technology in a manner that enables broad innovation and business improvement in the nuclear power plant operating model. Resolving long-term operational concerns with the II and C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation's energy and environmental security

Advanced Instrumentation, Information, and Control Systems Technologies Research in Support of Light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Hallbert, Bruce P.; Kenneth, Thomas [Idaho National Laboratory, Idaho (United States)

2014-08-15

The Advanced Instrumentation, Information, and Control (II and C) Systems Technologies Pathway conducts targeted research and development (R and D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals to ensure that legacy analog II and C systems are not life-limiting issues for the LWR fleet, and to implement digital II and C technology in a manner that enables broad innovation and business improvement in the nuclear power plant operating model. Resolving long-term operational concerns with the II and C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation's energy and environmental security.

78 FR 23935 - Federal Acquisition Regulation; Information Collection; Contractor Use of Interagency Fleet...

Science.gov (United States)

2013-04-23

...; Information Collection; Contractor Use of Interagency Fleet Management System Vehicles AGENCY: Department of... previously approved information collection requirement concerning contractor use of interagency fleet... Collection 9000- 0032, Contractor Use of Interagency Fleet Management System Vehicles, by any of the...

Three dimensional contact/impact methodology

International Nuclear Information System (INIS)

Kulak, R.F.

1987-01-01

The simulation of three-dimensional interface mechanics between reactor components and structures during static contact or dynamic impact is necessary to realistically evaluate their structural integrity to off-normal loads. In our studies of postulated core energy release events, we have found that significant structure-structure interactions occur in some reactor vessel head closure designs and that fluid-structure interactions occur within the reactor vessel. Other examples in which three-dimensional interface mechanics play an important role are: (1) impact response of shipping casks containing spent fuel, (2) whipping pipe impact on reinforced concrete panels or pipe-to-pipe impact after a pipe break, (3) aircraft crash on secondary containment structures, (4) missiles generated by turbine failures or tornados, and (5) drops of heavy components due to lifting accidents. The above is a partial list of reactor safety problems that require adequate treatment of interface mechanics and are discussed in this paper

An approach to the drone fleet survivability assessment based on a stochastic continues-time model

Science.gov (United States)

Kharchenko, Vyacheslav; Fesenko, Herman; Doukas, Nikos

2017-09-01

An approach and the algorithm to the drone fleet survivability assessment based on a stochastic continues-time model are proposed. The input data are the number of the drones, the drone fleet redundancy coefficient, the drone stability and restoration rate, the limit deviation from the norms of the drone fleet recovery, the drone fleet operational availability coefficient, the probability of the drone failure-free operation, time needed for performing the required tasks by the drone fleet. The ways for improving the recoverable drone fleet survivability taking into account amazing factors of system accident are suggested. Dependencies of the drone fleet survivability rate both on the drone stability and the number of the drones are analysed.

The heating operational summarization in three winters of a 5 MW test heating reactor

International Nuclear Information System (INIS)

Wang Dazhong; Dong Duo; Su Qingshan; Zhang Yajun

1992-09-01

The 5 MW THR (5 MW test heating reactor) is a new type reactor with inherent safety developed by INET (Institute of Nuclear Energy Technology). It is the first 'pressure vessel type' heating reactor in operation in the world. It was put into operation in November, 1989. Since then it has operated for three winter seasons. The total operation time has reached to 8174 hours and its availability of heating has reached to 99%. The advanced technology of this reactor has been proved in the past three years operation. The characteristics of power regulating, load following, reactivity disturbance and the variation of parameters under the condition of ATWS (anticipated transients without scram) were studied with experiments in 5 MW THR. The 5 MW THR is an ideal heating reactor and has outstanding performances

Materials for heat flux components of the first wall in fusion reactors

International Nuclear Information System (INIS)

Hoven, H.; Koizlik, K.; Linke, J.; Nickel, H.; Wallura, E.

1985-08-01

Materials of the First Wall in near-fusion plasma machines are subjected to a complex load system resulting from the plasma-wall interaction. The materials for their part also influence the plasma. Suitable materials must be available in order to ensure that the wall components achieve a sufficiently long dwell time and that their effects on the plasma remain small and controllable. The present report discusses relations between the plasma-wall interaction, the reactions of the materials and testing and examination methods for specific problems in developing and selecting suitable materials for highly stressed components on the First Wall of fusion reactors. (orig.)

RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

Energy Technology Data Exchange (ETDEWEB)

Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.

1996-02-01

In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components.

Application of a greedy algorithm to military aircraft fleet retirements

NARCIS (Netherlands)

Newcamp, J.M.; Verhagen, W.J.C.; Udluft, H.; Curran, Ricky

2017-01-01

This article presents a retirement analysis model for aircraft fleets. By employing a greedy algorithm, the presented solution is capable of identifying individually weak assets in a fleet of aircraft with inhomogeneous historical utilization. The model forecasts future retirement scenarios

Evaluation of Gear Condition Indicator Performance on Rotorcraft Fleet

Science.gov (United States)

Antolick, Lance J.; Branning, Jeremy S.; Wade, Daniel R.; Dempsey, Paula J.

2010-01-01

The U.S. Army is currently expanding its fleet of Health Usage Monitoring Systems (HUMS) equipped aircraft at significant rates, to now include over 1,000 rotorcraft. Two different on-board HUMS, the Honeywell Modern Signal Processing Unit (MSPU) and the Goodrich Integrated Vehicle Health Management System (IVHMS), are collecting vibration health data on aircraft that include the Apache, Blackhawk, Chinook, and Kiowa Warrior. The objective of this paper is to recommend the most effective gear condition indicators for fleet use based on both a theoretical foundation and field data. Gear diagnostics with better performance will be recommended based on both a theoretical foundation and results of in-fleet use. In order to evaluate the gear condition indicator performance on rotorcraft fleets, results of more than five years of health monitoring for gear faults in the entire HUMS equipped Army helicopter fleet will be presented. More than ten examples of gear faults indicated by the gear CI have been compiled and each reviewed for accuracy. False alarms indications will also be discussed. Performance data from test rigs and seeded fault tests will also be presented. The results of the fleet analysis will be discussed, and a performance metric assigned to each of the competing algorithms. Gear fault diagnostic algorithms that are compliant with ADS-79A will be recommended for future use and development. The performance of gear algorithms used in the commercial units and the effectiveness of the gear CI as a fault identifier will be assessed using the criteria outlined in the standards in ADS-79A-HDBK, an Army handbook that outlines the conversion from Reliability Centered Maintenance to the On-Condition status of Condition Based Maintenance.

Computation of the mechanical behaviour of nuclear reactor components

International Nuclear Information System (INIS)

Brosi, S.; Niffenegger, M.; Roesel, R.; Reichlin, K.; Duijvestijn, A.

1994-01-01

A possible limiting factor of the service life of a reactor is the mechanical load carrying margin, i.e. the excess of the load carrying capacity over the actual loading, of the central, heavy section components. This margin decreases during service but, for safety reasons, may not fall below a critical value. Therefore, it is essential to check and to control continuously the factors which cause the decrease. The reasons for the decrease are shown at length and in detail in an example relating to the test which almost achieved failure of a pipe emanating from a reactor pressure vessel, weakened by an artificial crack and undergoing a water-hammer loading. The latter was caused by a sudden valve closure supposed to follow upon a break far downstream. The computational and experimental difficulties associated with the simultaneous occurrence of an extreme weakening and an extreme loading in an already rather complicated geometry are explained. It is concluded that available computational tools and present know-how are sufficient to simulate the behaviour under such conditions as would prevail in normal service, and even to analyse departures from them, as long as not all difficulties arise simultaneously. (author) figs., tabs., refs

Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Ballinger, Ronald G. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Wang, Chun Yun [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Kadak, Andrew [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Todreas, Neil [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mirick, Bradley [Concepts, Northern Engineering and Research, Woburn, MA (United States); Demetri, Eli [Concepts, Northern Engineering and Research, Woburn, MA (United States); Koronowski, Martin [Concepts, Northern Engineering and Research, Woburn, MA (United States)

2004-08-30

The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a Generation IV nuclear system. The availability of controllable helium turbomachinery and compact heat exchangers are thus the critical enabling technology for the gas turbine cycle. The development of an initial reference design for an indirect helium cycle has been accomplished with the overriding constraint that this design could be built with existing technology and complies with all current codes and standards. Using the initial reference design, limiting features were identified. Finally, an optimized reference design was developed by identifying key advances in the technology that could reasonably be expected to be achieved with limited R&D. This final reference design is an indirect, intercooled and recuperated cycle consisting of a three-shaft arrangement for the turbomachinery system. A critical part of the design process involved the interaction between individual component design and overall plant performance. The helium cycle overall efficiency is significantly influenced by performance of individual components. Changes in the design of one component, a turbine for example, often required changes in other components. To allow for the optimization of the overall design with these interdependencies, a detailed steady state and transient control model was developed. The use of the steady state and transient models as a part of an iterative design process represents a key contribution of this work. A dynamic model, MPBRSim, has been developed. The model integrates the reactor core and the power conversion system simultaneously. Physical parameters such as the heat exchangers; weights and practical performance maps such as the turbine characteristics and compressor characteristics are incorporated into the model. The individual component models as well as the fully integrated model of the

Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

International Nuclear Information System (INIS)

Ballinger, Ronald G.; Chunyun Wang; Kadak, Andrew; Todreas, Neil

2004-01-01

The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a Generation IV nuclear system. The availability of controllable helium turbomachinery and compact heat exchangers are thus the critical enabling technology for the gas turbine cycle. The development of an initial reference design for an indirect helium cycle has been accomplished with the overriding constraint that this design could be built with existing technology and complies with all current codes and standards. Using the initial reference design, limiting features were identified. Finally, an optimized reference design was developed by identifying key advances in the technology that could reasonably be expected to be achieved with limited R and D. This final reference design is an indirect, intercooled and recuperated cycle consisting of a three-shaft arrangement for the turbomachinery system. A critical part of the design process involved the interaction between individual component design and overall plant performance. The helium cycle overall efficiency is significantly influenced by performance of individual components. Changes in the design of one component, a turbine for example, often required changes in other components. To allow for the optimization of the overall design with these interdependencies, a detailed steady state and transient control model was developed. The use of the steady state and transient models as a part of an iterative design process represents a key contribution of this work. A dynamic model, MPBRSim, has been developed. The model integrates the reactor core and the power conversion system simultaneously. Physical parameters such as the heat exchangers; weights and practical performance maps such as the turbine characteristics and compressor characteristics are incorporated into the model. The individual component models as well as the fully integrated model of the

Dynamic model of organic pollutant degradation in three dimensional packed bed electrode reactor.

Science.gov (United States)

Pang, Tianting; Wang, Yan; Yang, Hui; Wang, Tianlei; Cai, Wangfeng

2018-04-21

A dynamic model of semi-batch three-dimensional electrode reactor was established based on the limiting current density, Faraday's law, mass balance and a series of assumptions. Semi-batch experiments of phenol degradation were carried out in a three-dimensional electrode reactor packed with activated carbon under different conditions to verify the model. The factors such as the current density, the electrolyte concentration, the initial pH value, the flow rate of organic and the initial organic concentration were examined to know about the pollutant degradation in the three-dimensional electrode reactor. The various concentrations and logarithm of concentration of phenol with time were compared with the dynamic model. It was shown that the calculated data were in good agreement with experimental data in most cases. Copyright © 2018 Elsevier Ltd. All rights reserved.

DESIGNING OF AN AUTOMOBILE FLEET NETWORK

Directory of Open Access Journals (Sweden)

R. B. Ivut

2016-01-01

Full Text Available Investment volume is considered as an important factor in regional development under current conditions. Logistical infrastructure which ensures a complex transport, distributive, information and other services exerts a significant influence on regional investment attractiveness. Lack of clear vision on development and execution of development strategy for logistics infrastructure from the side of regional authorities results in unwillingness of large federal and transnational companies to provide investments in infrastructure projects. Network of automotive transport terminals is one of the main elements in logistics infrastructure. The network allows to optimize a flow of material goods from the point of their origin to the point of their consumption with the lowest possible costs and the required level of service. Automobile transport is one of the main objects of transport infrastructure and it is characterized by rather high flexibility in comparison with other types of transport facilities that preconditions its widespread application. Network of automobile fleets (terminals has been formed for redistribution of goods traffic within the concerned regions. The purpose of the present research is to develop a mathematical model for formation of transport infrastructure on the territory of regions. The paper proposes an approach for formation of automobile fleet (terminal network on the territory of a large region with due account of the established network of distribution and sorting-out warehouse facilities. A model has been developed for solving the problem pertaining to minimization of aggregate costs related to maintenance of automobile fleets, delivery of goods to and from distribution and sorting-out warehouse facilities to consumers, ferry of empty trucks and goods handling. The model makes it possible to determine optimal number and location area of automobile fleets (terminals while accounting for their possible locations, capacity

Development and application of a welding procedure for remote repair of Magnox reactor internal components

International Nuclear Information System (INIS)

Morgan-Warren, E.J.

1988-01-01

This paper summarises the development and application of an all-welding repair method for reinforcing magnox reactor internal components. The development was dominated by the necessity for remote operation and the environmental constraints, in particular the oxide covering on the steel reactor structure. The choice of welding process is described, together with the development of the procedure for remote operation. The quality assurance procedure, including the verification of the technique and monitoring of the repair operation, is discussed. (author)

48 CFR 51.204 - Use of interagency fleet management system (IFMS) vehicles and related services.

Science.gov (United States)

2010-10-01

... Contractor Use of Interagency Fleet Management System (IFMS) 51.204 Use of interagency fleet management system (IFMS) vehicles and related services. Contractors authorized to use interagency fleet management... 48 Federal Acquisition Regulations System 1 2010-10-01 2010-10-01 false Use of interagency fleet...

Application of data mining in three-dimensional space time reactor model

International Nuclear Information System (INIS)

Jiang Botao; Zhao Fuyu

2011-01-01

A high-fidelity three-dimensional space time nodal method has been developed to simulate the dynamics of the reactor core for real time simulation. This three-dimensional reactor core mathematical model can be composed of six sub-models, neutron kinetics model, cay heat model, fuel conduction model, thermal hydraulics model, lower plenum model, and core flow distribution model. During simulation of each sub-model some operation data will be produced and lots of valuable, important information reflecting the reactor core operation status could be hidden in, so how to discovery these information becomes the primary mission people concern. Under this background, data mining (DM) is just created and developed to solve this problem, no matter what engineering aspects or business fields. Generally speaking, data mining is a process of finding some useful and interested information from huge data pool. Support Vector Machine (SVM) is a new technique of data mining appeared in recent years, and SVR is a transformed method of SVM which is applied in regression cases. This paper presents only two significant sub-models of three-dimensional reactor core mathematical model, the nodal space time neutron kinetics model and the thermal hydraulics model, based on which the neutron flux and enthalpy distributions of the core are obtained by solving the three-dimensional nodal space time kinetics equations and energy equations for both single and two-phase flows respectively. Moreover, it describes that the three-dimensional reactor core model can also be used to calculate and determine the reactivity effects of the moderator temperature, boron concentration, fuel temperature, coolant void, xenon worth, samarium worth, control element positions (CEAs) and core burnup status. Besides these, the main mathematic theory of SVR is introduced briefly next, on the basis of which SVR is applied to dealing with the data generated by two sample calculation, rod ejection transient and axial

48 CFR 52.251-2 - Interagency Fleet Management System Vehicles and Related Services.

Science.gov (United States)

2010-10-01

... CLAUSES Text of Provisions and Clauses 52.251-2 Interagency Fleet Management System Vehicles and Related Services. As prescribed in 51.205, insert the following clause: Interagency Fleet Management System... to obtain interagency fleet management system vehicles and related services for use in the...

Application of the regulations on pressurized components or light water reactor primary coolant circuits

International Nuclear Information System (INIS)

Barthelemy, F.; Menjon, G.

1977-01-01

This paper describes the philosophy and the provisions of the Order of 26 February 1974 concerning application of the regulations on pressurized components for light water reactor steam supply systems. The aim is to show how these regulations which differ from other regulations on pressurized components and is more detailed on many points, is applied in practice in France in the various stages of the design, construction and operation of PWRs. (NEA) [fr

Assessment and management of ageing of major nuclear power plant components important to safety: CANDU reactor assemblies

International Nuclear Information System (INIS)

2001-02-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wearout of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring, and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs) including the Soviet designed water moderated and water cooled energy reactors (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age-related licensing issues. Since the reports are written from a safety perspective, they do not address life or life-cycle management of the plant components, which

CEMENT TRANSPORTATION LIMITED-FLEET MODELING AND ASSIGNING TO RATED DEMANDS

Directory of Open Access Journals (Sweden)

Narjes MASHHADI BANDANI

2017-04-01

Full Text Available Transportation is an inseparable part of the supply chain, with a key role in product distribution. This role is highlighted when ratio of “the cost of transportation” to “the value of goods” such as cement is significant. Iran has recently become one of the main centers of cement production in the world. However, transportation is the most important challenge in cement distribution because of weak structure of the transportation fleet and its independent action. Independence of and lack of commitment on the part of transportation fleets to cement companies as well as lack of timely delivery due to shortage of transportation in some routes and seasons lead to customers` dissatisfaction and even market loss or lack of market development. One of the significant differences between the transportation system in Iran and that in developed countries is lack of complete productivity of the transportation fleet. It means that trucks are driver-based in Iran. This paper introduces a model considering some issues such as driver-based trucks, size of the transportation fleet based on the number of active trucks, and demand priorities in the cement company. Taking the relation between the number of active trucks and the cement company into account, this model assigns weekly demands to the transportation fleet. It also tries to minimize the delay to respond to demands and increases the efficiency of the transportation fleet. Finally, this current condition-based model is compared with two other models including “no constraints on different routes of trucks” as well as single-route model for trucks.

Electric and Plug-In Hybrid Electric Fleet Vehicle Testing | Transportation

Science.gov (United States)

Research | NREL Electric and Plug-In Hybrid Electric Fleet Vehicle Evaluations Electric and Plug-In Hybrid Electric Fleet Vehicle Evaluations How Electric and Plug-In Hybrid Electric Vehicles plugging the vehicle into an electric power source. PHEVs are powered by an internal combustion engine that

Clean Cities Plug-In Electric Vehicle Handbook for Fleet Managers

Energy Technology Data Exchange (ETDEWEB)

None

2012-04-01

Plug-in electric vehicles (PEVs) are entering the automobile market and are viable alternatives to conventional vehicles. This guide for fleet managers describes the basics of PEV technology, PEV benefits for fleets, how to select the right PEV, charging a PEV, and PEV maintenance.

The Fleet-Management as an Element of the Modern Development in the Ukrainian Leasing Market

Directory of Open Access Journals (Sweden)

Garafonova Olga I.

2017-11-01

Full Text Available The article is aimed at researching practical aspects of the fleet-management development in Ukraine; relationships that arise in this process; elements that interact in the market. The preconditions of the companies’ transition to the use of this concept in the management of fleet, as well as the opportunities that arise in the enterprises using the principles of fleet-management were considered. The main advantages and disadvantages of the implementation of fleet-management in companies with the available fleet of various sizes were researched. The main features, current tendencies of the fleet-management development have been indicated, ways of attraction of automobile enterprises to the use of leasing have been suggested. Prospects for further researches in this direction will be consideration of possibility of attraction of more enterprises to the use of principles of fleet-management in their management routine. Further development of fleet management will provide enterprises with significant opportunities and the reserve funds that can be used for development.

Development of and verification test integral reactor major components - Development of manufacturing process and fabrication of prototype for SG and CEDM

Energy Technology Data Exchange (ETDEWEB)

Lee, Chang Hee; Park, Hwa Kyu; Kim, Yong Kyu; Choi, Yong Soon; Kang, Ki Su; Hyun, Young Min [Korea Heavy Industries and Construction Co., LTD., Changwon (Korea)

1999-03-01

Integral SMART(System integrated Modular Advanced Reactor) type reactor is under conceptual design. Because major components is integrated within in a single pressure vessel, compact design using advanced technology is essential. It means that manufacturing process for these components is more complex and difficult. The objective of this study is to confirm the possibility of manufacture of Steam Generator, Control Element Drive Mechanism(CEDM) and Reactor Assembly which includes Reactor Pressure Vessel, it is important to understand the design requirement and function of the major components. After understanding the design requirement and function, it is concluded that the helical bending and weld qualification of titanium tube for Steam Generator and the applicability of electron beam weld for CEDM step motor parts is the critical to fabricate the components. Therefore, bending mock-up and weld qualification of titanium tube was performed and the results are quite satisfactory. Also, it is concluded that electron beam welding technique can be applicable to the CEDM step motor part. (author). 22 refs., 14 figs., 46 tabs.

State of the art seismic analysis for CANDU reactor structure components using condensation method

Energy Technology Data Exchange (ETDEWEB)

Soliman, S A; Ibraham, A M; Hodgson, S [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

1996-12-31

The reactor structure assembly seismic analysis is a relatively complex process because of the intricate geometry with many different discontinuities, and due to the hydraulic attached mass which follows the structure during its vibration. In order to simulate reasonably accurate behaviour of the reactor structure assembly, detailed finite element models are generated and used for both modal and stress analysis. Guyan reduction condensation method was used in the analysis. The attached mass, which includes the fluid mass contained in the components plus the added mass which accounts for the inertia of the surrounding fluid entrained by the accelerating structure immersed in the fluid, was calculated and attached to the vibrating structures. The masses of the attached components, supported partly or totally by the assembly which includes piping, reactivity control units, end fittings, etc. are also considered in the analysis. (author). 4 refs., 6 tabs., 4 figs.

Optimizing a three-element core design for the Advanced Neutron Source Reactor

International Nuclear Information System (INIS)

West, C.D.

1995-01-01

Source of neutrons in the proposed Advanced Neutron Source facility is a multipurpose research reactor providing 5-10 times the flux, for neutron beams, of the best existing facilities. Baseline design for the reactor core, based on the ''no new inventions'' rule, was an assembly of two annular fuel elements similar to those used in the Oak Ridge and Grenoble high flux reactors, containing highly enriched U silicide particles. DOE commissioned a study of the use of medium- or low-enriched U; a three-element core design was studied as a means to provide extra volume to accommodate the additional U compound required when the fissionable 235 U has to be diluted with 238 U to reduce the enrichment. This paper describes the design and optimization of that three-element core

Fleet Planning Decision-Making: Two-Stage Optimization with Slot Purchase

Directory of Open Access Journals (Sweden)

Lay Eng Teoh

2016-01-01

Full Text Available Essentially, strategic fleet planning is vital for airlines to yield a higher profit margin while providing a desired service frequency to meet stochastic demand. In contrast to most studies that did not consider slot purchase which would affect the service frequency determination of airlines, this paper proposes a novel approach to solve the fleet planning problem subject to various operational constraints. A two-stage fleet planning model is formulated in which the first stage selects the individual operating route that requires slot purchase for network expansions while the second stage, in the form of probabilistic dynamic programming model, determines the quantity and type of aircraft (with the corresponding service frequency to meet the demand profitably. By analyzing an illustrative case study (with 38 international routes, the results show that the incorporation of slot purchase in fleet planning is beneficial to airlines in achieving economic and social sustainability. The developed model is practically viable for airlines not only to provide a better service quality (via a higher service frequency to meet more demand but also to obtain a higher revenue and profit margin, by making an optimal slot purchase and fleet planning decision throughout the long-term planning horizon.

ARCHER Project: Progress on Material and component activities for the Advanced High Temperature Reactor

International Nuclear Information System (INIS)

Buckthorpe, D.E.

2014-01-01

The ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R&D) integrated project is a four year project which was started in 2011 as part of the European Commission 7th Framework Programme (FP7) to perform High Temperature Reactor technology R&D in support of reactor demonstration. The project consortium encompasses conventional and Nuclear Industry, Utilities, Technical Support Organizations, Research & Development Organizations and Academia. The activities involved contribute to the Generation IV (GIF) International Forum and collaborate with related projects in the US, China, Japan, and the Republic of Korea in cooperation with IAEA and ISTC. This paper addresses the progress of the work on ARCHER materials and component activities since the start of the project and underlines some of the main conclusions reached. (author)

76 FR 1521 - Security Zone: Fleet Industrial Supply Center Pier, San Diego, CA

Science.gov (United States)

2011-01-11

...-AA87 Security Zone: Fleet Industrial Supply Center Pier, San Diego, CA AGENCY: Coast Guard, DHS. ACTION... Diego, CA. The existing security zone is around the former Fleet Industrial Supply Center Pier. The security zone encompasses all navigable waters within 100 feet of the former Fleet Industrial Supply Center...

The liquefied natural gas infrastructure and tanker fleet sizing problem

DEFF Research Database (Denmark)

Koza, David Franz; Røpke, Stefan; Molas, Anna Boleda

2017-01-01

We consider a strategic infrastructure and tanker fleet sizing problem in the liquefied natural gas business. The goal is to minimize long-term on-shore infrastructure and tanker investment cost combined with interrelated expected cost for operating the tanker fleet. A non-linear arc-based model...

Mathematical modeling of a three-phase trickle bed reactor

Directory of Open Access Journals (Sweden)

J. D. Silva

2012-09-01

Full Text Available The transient behavior in a three-phase trickle bed reactor system (N2/H2O-KCl/activated carbon, 298 K, 1.01 bar was evaluated using a dynamic tracer method. The system operated with liquid and gas phases flowing downward with constant gas flow Q G = 2.50 x 10-6 m³ s-1 and the liquid phase flow (Q L varying in the range from 4.25x10-6 m³ s-1 to 0.50x10-6 m³ s-1. The evolution of the KCl concentration in the aqueous liquid phase was measured at the outlet of the reactor in response to the concentration increase at reactor inlet. A mathematical model was formulated and the solutions of the equations fitted to the measured tracer concentrations. The order of magnitude of the axial dispersion, liquid-solid mass transfer and partial wetting efficiency coefficients were estimated based on a numerical optimization procedure where the initial values of these coefficients, obtained by empirical correlations, were modified by comparing experimental and calculated tracer concentrations. The final optimized values of the coefficients were calculated by the minimization of a quadratic objective function. Three correlations were proposed to estimate the parameters values under the conditions employed. By comparing experimental and predicted tracer concentration step evolutions under different operating conditions the model was validated.

How many reactor accidents will there be

International Nuclear Information System (INIS)

Islam, S.; Lindgren, K.

1986-01-01

A method for calculation of the probability of nuclear accidents is described. The method is based on the use of data from reactor operating experience, i.e. there have been two major accidents [Three Mile Island and Chernobyl] during 4,000 reactor-years (cumulative operating experience). The authors argue that this method is better than the present ''technical risk assessment'' method based on the likelihood of failure of a reactor component or safety system, used by designers of nuclear reactor. (U.K.)

Modeling and control of a large nuclear reactor. A three-time-scale approach

Energy Technology Data Exchange (ETDEWEB)

Shimjith, S.R. [Indian Institute of Technology Bombay, Mumbai (India); Bhabha Atomic Research Centre, Mumbai (India); Tiwari, A.P. [Bhabha Atomic Research Centre, Mumbai (India); Bandyopadhyay, B. [Indian Institute of Technology Bombay, Mumbai (India). IDP in Systems and Control Engineering

2013-07-01

Recent research on Modeling and Control of a Large Nuclear Reactor. Presents a three-time-scale approach. Written by leading experts in the field. Control analysis and design of large nuclear reactors requires a suitable mathematical model representing the steady state and dynamic behavior of the reactor with reasonable accuracy. This task is, however, quite challenging because of several complex dynamic phenomena existing in a reactor. Quite often, the models developed would be of prohibitively large order, non-linear and of complex structure not readily amenable for control studies. Moreover, the existence of simultaneously occurring dynamic variations at different speeds makes the mathematical model susceptible to numerical ill-conditioning, inhibiting direct application of standard control techniques. This monograph introduces a technique for mathematical modeling of large nuclear reactors in the framework of multi-point kinetics, to obtain a comparatively smaller order model in standard state space form thus overcoming these difficulties. It further brings in innovative methods for controller design for systems exhibiting multi-time-scale property, with emphasis on three-time-scale systems.

Path Planning Algorithms for the Adaptive Sensor Fleet

Science.gov (United States)

Stoneking, Eric; Hosler, Jeff

2005-01-01

The Adaptive Sensor Fleet (ASF) is a general purpose fleet management and planning system being developed by NASA in coordination with NOAA. The current mission of ASF is to provide the capability for autonomous cooperative survey and sampling of dynamic oceanographic phenomena such as current systems and algae blooms. Each ASF vessel is a software model that represents a real world platform that carries a variety of sensors. The OASIS platform will provide the first physical vessel, outfitted with the systems and payloads necessary to execute the oceanographic observations described in this paper. The ASF architecture is being designed for extensibility to accommodate heterogenous fleet elements, and is not limited to using the OASIS platform to acquire data. This paper describes the path planning algorithms developed for the acquisition phase of a typical ASF task. Given a polygonal target region to be surveyed, the region is subdivided according to the number of vessels in the fleet. The subdivision algorithm seeks a solution in which all subregions have equal area and minimum mean radius. Once the subregions are defined, a dynamic programming method is used to find a minimum-time path for each vessel from its initial position to its assigned region. This path plan includes the effects of water currents as well as avoidance of known obstacles. A fleet-level planning algorithm then shuffles the individual vessel assignments to find the overall solution which puts all vessels in their assigned regions in the minimum time. This shuffle algorithm may be described as a process of elimination on the sorted list of permutations of a cost matrix. All these path planning algorithms are facilitated by discretizing the region of interest onto a hexagonal tiling.

3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

Energy Technology Data Exchange (ETDEWEB)

Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra R, G.; Del Valle G, E.; Sanchez R, A., E-mail: narehc@hotmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)

2016-09-15

The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

41 CFR 102-34.340 - Do we need a fleet management information system?

Science.gov (United States)

2010-07-01

... management information system? 102-34.340 Section 102-34.340 Public Contracts and Property Management Federal... VEHICLE MANAGEMENT Federal Fleet Report § 102-34.340 Do we need a fleet management information system? Yes, you must have a fleet management information system at the department or agency level that — (a...

Feasible reactor power cutback logic development for an integral reactor

International Nuclear Information System (INIS)

Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

2013-01-01

Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

Application of environmentally-corrected fatigue curves to nuclear power plant components

International Nuclear Information System (INIS)

Ware, A.G.; Morton, D.K.; Nitzel, M.E.

1996-01-01

Recent test data indicate that the effects of the light water reactor (LWR) environment could significantly reduce the fatigue resistance of materials used in the reactor coolant pressure boundary components of operating nuclear power plants. Argonne National Laboratory has developed interim fatigue curves based on test data simulating LWR conditions, and published them in NUREG/CR-5999. In order to assess the significance of these interim fatigue curves, fatigue evaluations of a sample of the components in the reactor coolant pressure boundary of LWRs were performed. The sample consists of components from facilities designed by each of the four US nuclear steam supply system vendors. For each facility, six locations were studied including two locations on the reactor pressure vessel. In addition, there are older vintage plants where components of the reactor coolant pressure boundary were designed to codes that did not require an explicit fatigue analysis of the components. In order to assess the fatigue resistance of the older vintage plants, an evaluation was also conducted on selected components of three of these plants. This paper discusses the insights gained from the application of the interim fatigue curves to components of seven operating nuclear power plants

Uranium in phosphate rocks and future nuclear power fleets

International Nuclear Information System (INIS)

Gabriel, S.; Baschwitz, A.; Mathonniere, G.

2014-01-01

phosphoric acid. Uranium recovery as a by-product of phosphate rocks could be competitive for the moment, but limited at the most to 10 kt U per year, i.e. less than 20% of current world demand. The only way to lift the constraint of capacity production is to produce uranium as a primary product of phosphates. Unfortunately, this solution is very unlikely due to its high unit cost. In line with these considerations, the correspondence between the estimated resources and the forecast energy scenarios is examined, first with the current type of light water reactors which burn uranium, and secondly with a mixed fleet with both light water reactors and fast reactors which use plutonium. (author)

A multi-agent based intelligent configuration method for aircraft fleet maintenance personnel

Directory of Open Access Journals (Sweden)

Feng Qiang

2014-04-01

Full Text Available A multi-agent based fleet maintenance personnel configuration method is proposed to solve the mission oriented aircraft fleet maintenance personnel configuration problem. The maintenance process of an aircraft fleet is analyzed first. In the process each aircraft contains multiple parts, and different parts are repaired by personnel with different majors and levels. The factors and their relationship involved in the process of maintenance are analyzed and discussed. Then the whole maintenance process is described as a 3-layer multi-agent system (MAS model. A communication and reasoning strategy among the agents is put forward. A fleet maintenance personnel configuration algorithm is proposed based on contract net protocol (CNP. Finally, a fleet of 10 aircraft is studied for verification purposes. A mission type with 3 waves of continuous dispatch is imaged. Compared with the traditional methods that can just provide configuration results, the proposed method can provide optimal maintenance strategies as well.

PROBLEMS OF FORMATION OF RUSSIAN INNOVATIVE AGRICULTURAL TRACTORS FLEET

Directory of Open Access Journals (Sweden)

V. M. Kryazhkov

2015-01-01

Full Text Available Formation of tractors park and its perspective development are carried out on the basis of realization and improvement of the scientifically based types range constructed on a traction classification sign which conforms to requirements of agricultural production when performing two main conditions: full meeting the requirements for various standard sizes of tractors applicable to conditions of the corresponding zones of Russia and continuous reproduction and development of tractor fleet, both by quantity of efficient machinery and its structure. The developed types range practically runs the gamut of organizational and economic and climatic conditions, and also all range of forms of the organizations using machinery, beginning from individual consumers, small and large rent and contract collectives and finishing large-scale enterprises and their associations. Federal Service of State Statistics (Rosstat information about a steady tendency of fleet reduction was used for studying of regularities of development of tractor fleet. This information was added by analytical synthesis of data about the annual brand purchases in the Russian market of agricultural tractors. The developed database shows a share of the tractors in fleets of agricultural organizations, which are used over depreciation term, and also minimum necessary number of new machines introduction for formation of volume of tractors with limit age and the total quantity of tractors necessary for minimization of set of expenses. It was established that it is possible to estimate visually the developed quantitative and age structure of tractors fleet and to predict its development on prospect by worked out database representing set of annual tractors brand lines purchasing.

Nuclear reactor component populations, reliability data bases, and their relationship to failure rate estimation and uncertainty analysis

International Nuclear Information System (INIS)

Martz, H.F.; Beckman, R.J.

1981-12-01

Probabilistic risk analyses are used to assess the risks inherent in the operation of existing and proposed nuclear power reactors. In performing such risk analyses the failure rates of various components which are used in a variety of reactor systems must be estimated. These failure rate estimates serve as input to fault trees and event trees used in the analyses. Component failure rate estimation is often based on relevant field failure data from different reliability data sources such as LERs, NPRDS, and the In-Plant Data Program. Various statistical data analysis and estimation methods have been proposed over the years to provide the required estimates of the component failure rates. This report discusses the basis and extent to which statistical methods can be used to obtain component failure rate estimates. The report is expository in nature and focuses on the general philosophical basis for such statistical methods. Various terms and concepts are defined and illustrated by means of numerous simple examples

Preliminary assessment of fleets covered by the Energy Policy Act

Energy Technology Data Exchange (ETDEWEB)

Hu, P.S.; Davis, S.C.; Wang, M.Q. [and others

1994-12-31

To facilitate the goal of decreasing oil imports by 10 percent by the year 2000 and 30 percent by 2010, two sections of the Energy Policy Act encourage and mandate alternative fuel vehicles in the acquisition of fleet vehicles. The first step in estimating the contribution of these mandates toward meeting the aforementioned goal entails identifying affected fleets. This paper presents a preliminary assessment of potential vehicle fleet coverage. Only a limited number of companies in the methanol, ethanol, and hydrogen industries are likely to quality for this mandate. Whereas, many of the oil producers, petroleum refiners, and electricity companies are likely to be regulated.

A concise guide to effective management of the public works equipment fleet

OpenAIRE

De Armas, Vicente.

1996-01-01

CIVINS (Civilian Institutions) Thesis document Approved for public release ; distribution is unlimited The purpose of this report is to present in a concise form the principles and fundamentals of managing the public works equipment fleet, providing a basis from which to develop an effective fleet management program. The report begins with a review of today's trends in fleet management, including, organization, privatization, effects of environmental regulations and a discussion on info...

A multi-agent based intelligent configuration method for aircraft fleet maintenance personnel

OpenAIRE

Feng, Qiang; Li, Songjie; Sun, Bo

2014-01-01

A multi-agent based fleet maintenance personnel configuration method is proposed to solve the mission oriented aircraft fleet maintenance personnel configuration problem. The maintenance process of an aircraft fleet is analyzed first. In the process each aircraft contains multiple parts, and different parts are repaired by personnel with different majors and levels. The factors and their relationship involved in the process of maintenance are analyzed and discussed. Then the whole maintenance...

Modernization of reactor instrumentation for research reactors at Trombay

International Nuclear Information System (INIS)

Darbhe, M.D.; Chaudhuri, H.

1989-01-01

The three research reactors at Trombay, viz., Apsara, Cirus and Zerlina were commissioned in 1956, 1960 and 1961 respectively. The nuclear instrumentation designs were based on the vacuum tube technology, which was prevalent during those days. The effect of component obsolescence of critical components like vacuum tubes, magnetic amplifiers and sensitrol meter relays was strongly felt since early 1970s. Also, the failure rates of the units were observed to show an increasing trend due to ageing and lack of good quality indigenous spares. Hence it was proposed to replace the nuclear instrumentation units for the three reactors, with those employing modern, state of the art solid state devices, keeping indigenous content as high as practicable. The work started in 1977 with the preparations of specifications and the project was scheduled to be completed in 1981. The project was divided into two phases. The Phase I comprising of nuclear channels common to all reactors and Phase II consisting exclusively of regulating system units of Cirus. The salient stages of project progress and completion were: (i) Fabrication and testing of final design prototypes was completed by end of 1982. (ii) Commissioning of new units at Apsara was completed in January 1984. (iii) Commissioning of new units at Cirus was completed in September 1984. An account of experience in all these stages and problems encountered is given. (author). 6 figs

Core followup studies of the Tarapur Reactors with the three dimensional BWR simulator COMTEG

Energy Technology Data Exchange (ETDEWEB)

Dwivedi, S. R.; Jagannathan, V.; Mohanakrishnan, P.; Srinivasan, K. R.; Rastogi, B. P.

1976-07-01

Both the units of the Tarapur Atomic Power Station started operation in the year 1969. Since then, these units have completed three cycles. For efficient operation and fuel management of these reactors, a three dimensional BWR simulator COMETG has been developed. The reactors are closely being followed using the simulator. The detailed analyses for cycle 3/4 operation of both the units are described in the paper. The results show very good agreement between calculated and measured values. It is concluded that reactor core behaviour could be predicted in a satisfactory manner with the core simulator COMETG.

Impact of the Three Mile Island accident on reactor safety and licensing in Canada

International Nuclear Information System (INIS)

Harvie, J.D.

1980-06-01

This paper discusses the implications of the accident at Three Mile Island on reactor safety and licensing in Canada. Reactor safety principles which can be learned from, or are reaffirmed by, the accident are reviewed. It is concluded that reactor safety demands a firm commitment to safety by all those involved in the nuclear industry. (auth)

An ecologic accompanying research on the fleet test electromobility. Final report; Oekologische Begleitforschung zum Flottenversuch Elektromobilitaet. Endbericht

Energy Technology Data Exchange (ETDEWEB)

Helms, Heinrich; Lambrecht, Udo; Joehrens, Julius; Pehnt, Martin; Liebich, Axel; Weiss, Uta; Kaemper, Claudia

2013-06-15

Mobility is a prerequisite for numerous economic and private activities and thus a central component of our life. The demand for mobility in Germany is predominantly covered by the road traffic. The Institute for Energy and Environmental Research (Heidelberg, Federal Republic of Germany) carried out an ecologic accompanying research in the course of a fleet test in order to determine the potentials of environmental improvement of the vehicles tested in a field test. These potentials were perpetuated for future serial vehicles and projected to Germany. The main themes of this contribution are: (a) Production and disposal of plug-in-hybrids; (b) The TwinDrive in the fleet test; (c) Plug-in-hybrid serial vehicles 2020; (d) Perspective of electromobility in Germany 2030; (e) Strategic evaluation of the electomobility.

A systems approach to improving fleet policy compliance within the US Federal Government

International Nuclear Information System (INIS)

Deason, Kristin S.; Jefferson, Theresa

2010-01-01

To reduce dependence on foreign sources of energy, address climate change, and improve environmental quality, the US government has established a goal of reducing petroleum fuel use in its federal agencies. To this end, the government requires its agencies to purchase alternative fuel vehicles, use alternative fuel, and adopt other strategies to reduce petroleum consumption. Compliance with these requirements, while important, creates challenges for federal fleet managers who oversee large, geographically dispersed fleets. In this study, a group of 25 experienced federal fleet managers participated in a pilot study using a structured methodology for developing strategies to comply with fleet requirements while using agency resources as efficiently as possible. Multi-criteria decision making (MCDM) methods were used to identify and quantify agency priorities in combination with a linear programming model to optimize the purchase of fleet vehicles. The method was successful in quantifying tradeoffs and decreasing the amount of time required to develop fleet management strategies. As such, it is recommended to federal agencies as a standard tool for the development of these strategies in the future. (author)

A systems approach to improving fleet policy compliance within the US Federal Government

Energy Technology Data Exchange (ETDEWEB)

Deason, Kristin S. [The George Washington University, 1776 G St. NW, Washington, DC 20052 (United States); Jefferson, Theresa [Virginia Polytechnic Institute and State University, 1101 King St, Suite 610 Alexandria, VA 22314 (United States)

2010-06-15

To reduce dependence on foreign sources of energy, address climate change, and improve environmental quality, the US government has established a goal of reducing petroleum fuel use in its federal agencies. To this end, the government requires its agencies to purchase alternative fuel vehicles, use alternative fuel, and adopt other strategies to reduce petroleum consumption. Compliance with these requirements, while important, creates challenges for federal fleet managers who oversee large, geographically dispersed fleets. In this study, a group of 25 experienced federal fleet managers participated in a pilot study using a structured methodology for developing strategies to comply with fleet requirements while using agency resources as efficiently as possible. Multi-criteria decision making (MCDM) methods were used to identify and quantify agency priorities in combination with a linear programming model to optimize the purchase of fleet vehicles. The method was successful in quantifying tradeoffs and decreasing the amount of time required to develop fleet management strategies. As such, it is recommended to federal agencies as a standard tool for the development of these strategies in the future. (author)

Relationships between chemical oxygen demand (COD) components and toxicity in a sequential anaerobic baffled reactor/aerobic completely stirred reactor system treating Kemicetine

Energy Technology Data Exchange (ETDEWEB)

Sponza, Delia Teresa, E-mail: delya.sponza@deu.edu.tr [Department of Environmental Engineering, Faculty of Engineering, Dokuz Eyluel University, Buca Kaynaklar Campus, Tinaztepe, 35160 Izmir (Turkey); Demirden, Pinar, E-mail: pinar.demirden@kozagold.com [Environmental Engineer, Koza Gold Company, Environmental Department, Ovacik, Bergama Izmir (Turkey)

2010-04-15

In this study the interactions between toxicity removals and Kemicetine, COD removals, intermediate products of Kemicetine and COD components (CODs originating from slowly degradable organics, readily degradable organics, inert microbial products and from the inert compounds) were investigated in a sequential anaerobic baffled reactor (ABR)/aerobic completely stirred tank reactor (CSTR) system with a real pharmaceutical wastewater. The total COD and Kemicetine removal efficiencies were 98% and 100%, respectively, in the sequential ABR/CSTR systems. 2-Amino-1 (p-nitrophenil)-1,3 propanediol, l-p-amino phenyl, p-amino phenol and phenol were detected in the ABR as the main readily degradable inter-metabolites. In the anaerobic ABR reactor, the Kemicetin was converted to corresponding inter-metabolites and a substantial part of the COD was removed. In the aerobic CSTR reactor the inter-metabolites produced in the anaerobic reactor were completely removed and the COD remaining from the anerobic reactor was biodegraded. It was found that the COD originating from the readily degradable organics did not limit the anaerobic degradation process, while the CODs originating from the slowly degradable organics and from the inert microbial products significantly decreased the anaerobic ABR reactor performance. The acute toxicity test results indicated that the toxicity decreased from the influent to the effluent of the aerobic CSTR reactor. The ANOVA test statistics showed that there was a strong linear correlation between acute toxicity, CODs originating from the slowly degradable organics and inert microbial products. A weak correlation between acute toxicity and CODs originating from the inert compounds was detected.

Relationships between chemical oxygen demand (COD) components and toxicity in a sequential anaerobic baffled reactor/aerobic completely stirred reactor system treating Kemicetine

International Nuclear Information System (INIS)

Sponza, Delia Teresa; Demirden, Pinar

2010-01-01

In this study the interactions between toxicity removals and Kemicetine, COD removals, intermediate products of Kemicetine and COD components (CODs originating from slowly degradable organics, readily degradable organics, inert microbial products and from the inert compounds) were investigated in a sequential anaerobic baffled reactor (ABR)/aerobic completely stirred tank reactor (CSTR) system with a real pharmaceutical wastewater. The total COD and Kemicetine removal efficiencies were 98% and 100%, respectively, in the sequential ABR/CSTR systems. 2-Amino-1 (p-nitrophenil)-1,3 propanediol, l-p-amino phenyl, p-amino phenol and phenol were detected in the ABR as the main readily degradable inter-metabolites. In the anaerobic ABR reactor, the Kemicetin was converted to corresponding inter-metabolites and a substantial part of the COD was removed. In the aerobic CSTR reactor the inter-metabolites produced in the anaerobic reactor were completely removed and the COD remaining from the anerobic reactor was biodegraded. It was found that the COD originating from the readily degradable organics did not limit the anaerobic degradation process, while the CODs originating from the slowly degradable organics and from the inert microbial products significantly decreased the anaerobic ABR reactor performance. The acute toxicity test results indicated that the toxicity decreased from the influent to the effluent of the aerobic CSTR reactor. The ANOVA test statistics showed that there was a strong linear correlation between acute toxicity, CODs originating from the slowly degradable organics and inert microbial products. A weak correlation between acute toxicity and CODs originating from the inert compounds was detected.

Nuclear fleet: Today and tomorrow

International Nuclear Information System (INIS)

Levin, B.M.; Kovalenko, V.K.; Sinyaev, A.K.

1993-01-01

Many years of operational experience have shown advantages of nuclear liner icebreakers over those burning fossil fuels primarily in ensuring reliable and stable shipping along the North Sea Route with an extended navigation season being realized. The advantages of the nuclear fleet are described

Review of light water reactor safety through the Three Mile Island accident

International Nuclear Information System (INIS)

Phung, D.L.

1984-05-01

This review of light water reactor safety through the Three Mile Island accident has the purpose of establishing the baseline over which safety achievement post-TMI is assessed, and the need for new reactor designs and business direction is judged. Five major areas of reactor safety pre-TMI are examined: (1) safety philosophy and institutions, (2) reactor design criteria, (3) operational problems, (4) the Rasmussen reactor safety study, and (5) the TMI accident and repercussions. Although nuclear power has made spectacular achievements over the period pre-TMI and although TMI is technically a minor accident, this review concludes that there were basic flaws in the technology and in the manner safety philosophy was conceived and carried out. These flaws included (1) a reactor design that has high core power density, low heat capacity, and low system tolerance to upsets, (2) reactor deployment that had been expedited without extensive operational experience, (3) rules and regulations that had to play catch-up with commercial reactor development, (4) an industry that was fragmented, short-sighted, and tended to rely on the Nuclear Regulatory Commission for safety guidance, (5) information that was not effectively shared, and (6) attention that was inadequate to the human aspects of reactor operation and to public reaction to the specter of a reactor accident, major or minor

Optimization Model and Algorithm Design for Airline Fleet Planning in a Multiairline Competitive Environment

Directory of Open Access Journals (Sweden)

Yu Wang

2015-01-01

Full Text Available This paper presents a multiobjective mathematical programming model to optimize airline fleet size and structure with consideration of several critical factors severely affecting the fleet planning process. The main purpose of this paper is to reveal how multiairline competitive behaviors impact airline fleet size and structure by enhancing the existing route-based fleet planning model with consideration of the interaction between market share and flight frequency and also by applying the concept of equilibrium optimum to design heuristic algorithm for solving the model. Through case study and comparison, the heuristic algorithm is proved to be effective. By using the algorithm presented in this paper, the fleet operational profit is significantly increased compared with the use of the existing route-based model. Sensitivity analysis suggests that the fleet size and structure are more sensitive to the increase of fare price than to the increase of passenger demand.

Overview moderator material for nuclear reactor components

International Nuclear Information System (INIS)

Mairing Manutu Pongtuluran; Hendra Prihatnadi

2009-01-01

In order for a reactor design is considered acceptable absolute technical requirement is fulfilled because the most important part of a reactor design. Safety considerations emphasis on the handling of radioactive substances emitted during the operation of a reactor and radioactive waste handling. Moderator material is a layer that interacts directly with neutrons split the nuclear fuel that will lead to changes in physical properties, nuclear properties, mechanical properties and chemical properties. Reviews moderator of this time is of the types of moderator is often used to meet the requirements as nuclear material. (author)

A trial of retrofitted advisory collision avoidance technology in government fleet vehicles.

Science.gov (United States)

Thompson, James P; Mackenzie, Jamie R R; Dutschke, Jeffrey K; Baldock, Matthew R J; Raftery, Simon J; Wall, John

2018-06-01

In-vehicle collision avoidance technology (CAT) has the potential to prevent crash involvement. In 2015, Transport for New South Wales undertook a trial of a Mobileye 560 CAT system that was installed in 34 government fleet vehicles for a period of seven months. The system provided headway monitoring, lane departure, forward collision and pedestrian collision warnings, using audio and visual alerts. The purpose of the trial was to determine whether the technology could change the driving behaviour of fleet vehicle drivers and improve their safety. The evaluation consisted of three components: (1) analysis of objective data to examine effects of the technology on driving behaviour, (2) analysis of video footage taken from a sample of the vehicles to examine driving circumstances that trigger headway monitoring and forward collision warnings, and (3) a survey completed by 122 of the 199 individuals who drove the trial vehicles to examine experiences with, and attitudes to, the technology. Analysis of the objective data found that the system resulted in changes in behaviour with increased headway and improved lane keeping, but that these improvements dissipated once the warning alerts were switched off. Therefore, the system is capable of altering behaviour but only when it is actively providing alerts. In-vehicle video footage revealed that over a quarter of forward collision warnings were false alarms, in which a warning event was triggered despite there being no vehicle travelling ahead. The surveyed drivers recognised that the system could improve safety but most did not wish to use it themselves as they found it to be distracting and felt that it would not prevent them from having a crash. The results demonstrate that collision avoidance technology can improve driving behaviour but drivers may need to be educated about the potential benefits for their driving in order to accept the technology. Copyright © 2018 Elsevier Ltd. All rights reserved.

Structural integrity and management of aging in internal components of BWR reactors; Integridad estructural y manejo del envejecimiento en componentes internos de reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Arganis J, C.R. [Instituto Nacional de Investigaciones Nucleares, Km 36.5 Carretera Mexico, Toluca Salazar Edo. de Mexico (Mexico)]. E-mail: craj@nuclear.inin.mx

2004-07-01

Presently work the bases to apply structural integrity and the handling of the aging of internal components of the pressure vessel of boiling water reactors of water are revised and is carried out an example of structural integrity in the horizontal welding H4 of the encircling one of the core of a reactor, taking data reported in the literature. It is also revised what is required to carry out the handling program or conduct of the aging (AMP). (Author)

Impact of environmental constraints and aircraft technology on airline fleet composition

Science.gov (United States)

Moolchandani, Kushal A.

This thesis models an airline's decisions about fleet evolution in order to maintain economic and regulatory viability. The aim is to analyze the fleet evolution under different scenarios of environmental policy and technology availability in order to suggest an optimal fleet under each case. An understanding of the effect of aircraft technologies, fleet size and age distribution, and operational procedures on airline performance may improve the quality of policies to achieve environmental goals. Additionally, the effect of decisions about fleet evolution on air travel is assessed as the change in market demand and profits of an abstracted, benevolent monopolist airline. Attention to the environmental impact of aviation has grown, and this has prompted several organizations such as ICAO (and, in response, NASA) to establish emissions reduction targets to reduce aviation's global climate impact. The introduction of new technology, change in operational procedures, etc. are some of the proposed means to achieve these targets. Of these, this thesis studies the efficacy of implementation of environmental policies in form of emissions constraints as a means to achieve these goals and assesses their impact on an airline's fleet evolution and technology use (along with resulting effects on air travel demand). All studies in this thesis are conducted using the Fleet-level Environmental Evaluation Tool (FLEET), a NASA sponsored simulation tool developed at Purdue University. This tool models airline operational decisions via a resource allocation problem and uses a system dynamics type approach to mimic airline economics, their decisions regarding retirement and acquisition of aircraft and evolution of market demand in response to the economic conditions. The development of an aircraft acquisition model for FLEET is a significant contribution of the author. Further, the author conducted a study of various environmental policies using FLEET. Studies introduce constraints on

Three-dimensional reactor model for the Paks NPP full-scope simulator

International Nuclear Information System (INIS)

Gyori, C.; Hegyi, G.; Hozer, Z.; Kereszturi, A.; Maraczy, C.

1993-01-01

The reactor model includes thermohydraulic and neutron-physical components. The thermohydraulic model is based on the SMABRE code developed at the Technical Research Centre of Finland for the analysis of loss-of-coolant transients in PWRs. The fuel rod model will be replaced by a new software module providing a comprehensive description of the behavior of fuel rods during reactor transients and hypothetical accidents. The calculation is performed in four individual models: fuel rod temperature model, fuel rod internal pressure model, fuel rod deformation model and fuel rod failure model. In the neutron-physical model the core is calculated with nodes for all of the 349 fuel assemblies, and each assembly is calculated in ten layers. (Z.S.) 1 fig., 5 refs

Description of reactor fuel breeding with three integral concepts

International Nuclear Information System (INIS)

Ott, K.O.; Hanan, N.A.; Maudlin, P.J.; Borg, R.C.

1979-01-01

The time-dependent breeding of fuel in a growing system of breeder reactors can be characterized by the transitory (instantaneous) growth rate, γ(t). The three most important aspects of γ(t) can be expressed by time-independent integral concepts. Two of these concepts are in widespread use. A third integral concept that links the two earlier ones is introduced. The time-dependent growth rate has an asymptotic value, γ/sup infinity/, the equilibrium growth rate, which is the basis for the calculation of the doubling time. The equilibrium growth rate measures the breeding capability and represents a reactor property. Maximum deviation of γ(t) and γ/sup infinity/ generally appears at the initial startup of the reactor, where γ(t = 0) = γ 0 . This deviation is due to the difference between the initial and asymptotic fuel inventory composition. The initial growth rate can be considered a second integral concept; it characterizes the breeding of a particular fuel in a given reactor. Growth rates are logarithmic derivatives of the growing mass of fuel in breeder reactors, especially γ/sup infinity/, which describes the asymptotic growth by exp(γ/sup infinity/t). There is, however, a variation in the fuel-mass factor in front of this exponential function during the transition from γ 0 to γ/sup infinity/. It is shown that this variation of the fuel mass during transitioncan be described by a third integral concept, termed the breeding bonus, b. The breeding bonus measures the quality of a fuel for its use in a given reactor in terms of its impact on the magnitude of the asymptotically growing fuel mass. The calculation of γ 0 and γ/sup infinity/ is facilitated by use of the critical mass (CM) worths and the breeding worth factors, respectively

Phytoplankton distribution in three thermally different but edaphically similar reactor cooling reservoirs

Energy Technology Data Exchange (ETDEWEB)

Wilde, E W

1982-01-01

Phytoplankton community structure and the physicochemical characteristics of three reactor cooling reservoirs in close proximity and of similar age and bottom type were studied during 1978. The three reservoirs differed in thermal alteration resulting from reactor cooling water as follows: (1) considerable heating with lake-wide temperatures >30/sup 0/C, even in winter; (2) a maximal 5/sup 0/C increase occurring in only one of three major arms of the reservoir; and (3) no thermal effluent received during the study period. Considerable spatial and temporal differences in water quality and phytoplankton community structure were observed; however, water temperature independent of other environmental factors (e.g., light and nutrients) was found to be a relatively unimportant variable for explaining phytoplankton periodicity.

Phytoplankton distribution in three thermally different but edaphically similar reactor cooling reservoirs

International Nuclear Information System (INIS)

Wilde, E.W.

1982-01-01

Phytoplankton community structure and the physicochemical characteristics of three reactor cooling reservoirs in close proximity and of similar age and bottom type were studied during 1978. The three reservoirs differed in thermal alteration resulting from reactor cooling water as follows: (1) considerable heating with lake-wide temperatures >30 0 C, even in winter; (2) a maximal 5 0 C increase occurring in only one of three major arms of the reservoir; and (3) no thermal effluent received during the study period. Considerable spatial and temporal differences in water quality and phytoplankton community structure were observed; however, water temperature independent of other environmental factors (e.g., light and nutrients) was found to be a relatively unimportant variable for explaining phytoplankton periodicity

AVTA Federal Fleet PEV Readiness Data Logging and Characterization Study: Final Report

Energy Technology Data Exchange (ETDEWEB)

Schey, Stephen [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Francfort, Jim [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

2015-06-01

Collect and evaluate data on federal fleet operations as part of the Advanced Vehicle Testing Activity’s Federal Fleet Vehicle Data Logging and Characterization Study. The Advanced Vehicle Testing Activity study seeks to collect and evaluate data to validate the utilization of advanced plug-in electric vehicle (PEV) transportation. This report summarizes the fleets studied to identify daily operational characteristics of select vehicles and report findings on vehicle and mission characterizations to support the successful introduction of PEVs into the agencies’ fleets. Individual observations of these selected vehicles provide the basis for recommendations related to electric vehicle adoption and whether a battery electric vehicle or plug-in hybrid electric vehicle (collectively referred to as PEVs) can fulfill the mission requirements.

The Molten Salt Fast Reactor as Highly Efficient Transmutation System

International Nuclear Information System (INIS)

Merk, B.; Rohde, U.; Scholl, S.

2013-01-01

Conclusion and future steps: • MSFR offers very attractive features for efficient transmutation; • significant advantages due to liquid fuel and online refuelling and reprocessing; • significant developments are required on the way to application; • system is very promising for transmutation; • development of a safety approach for liquid fuel reactors (RSWG); • investigation of possibilities to solve the “last transmuter” problem (ICAPP2013) – as future for countries envisaging nuclear phase out or no transition to fast reactor fleet for energy production; • establishing of a strong group “MSFR for transmutation”; • development of a transmutation optimized design

Investigation of the local component of power-reactor noise via diffusion theory

International Nuclear Information System (INIS)

Kosaly, G.

1975-03-01

The aim of the paper is to provide a theoretical background for the phenomenological model, which postulates the existence of a local component in the neutron noise of a light water cooled boiling water reactor. After the introductory review of the phenomenological model, noise calculation are performed by help of the one-group and two-group diffusion theory. Only in the two-group diffusion model it is succeeded to find a term in the response to a propagating disturbance of density which results in a small volume of neutrons physical sensivity around the point of observation. The problem, whether this local component can be a dominating term in the solution or not, is investigated in the Appenix. (Sz.Z.)

Economic Analysis of Symbiotic Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)

International Nuclear Information System (INIS)

Williams, Kent Alan; Shropshire, David E.

2009-01-01

A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle

41 CFR 101-39.105-1 - Transfers from discontinued or curtailed fleet management systems.

Science.gov (United States)

2010-07-01

... discontinued or curtailed fleet management systems. 101-39.105-1 Section 101-39.105-1 Public Contracts and... AVIATION, TRANSPORTATION, AND MOTOR VEHICLES 39-INTERAGENCY FLEET MANAGEMENT SYSTEMS 39.1-Establishment, Modification, and Discontinuance of Interagency Fleet Management Systems § 101-39.105-1 Transfers from...

RB research nuclear reactor, Annual report for 1989, I - III; Istrazivacki nukleani reaktor RB (Izvestaj o radu u 1989. godini), I - III

Energy Technology Data Exchange (ETDEWEB)

Stefanovic, D; Pesic, M; Hadimahmutovic, N; Vranic, S; Petronijevic, M; Jevremovic, M; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1989-12-15

This report is made of three parts. Part one contains a short description of the reactor, reactor operation, incidents, status of reactor equipment and components (nuclear fuel, heavy water, reactor vessel, heavy water circulation system, electronic, electric and mechanical equipment, auxiliary systems and Vax-8250 computer). It includes dosimetry and radiation protection data, personnel and financial data. Second part of this report in concerned with maintenance of reactor components and instrumentation. Part three includes data about reactor utilization during 1989.

Fleet Sizing of Automated Material Handling Using Simulation Approach

Science.gov (United States)

Wibisono, Radinal; Ai, The Jin; Ratna Yuniartha, Deny

2018-03-01

Automated material handling tends to be chosen rather than using human power in material handling activity for production floor in manufacturing company. One critical issue in implementing automated material handling is designing phase to ensure that material handling activity more efficient in term of cost spending. Fleet sizing become one of the topic in designing phase. In this research, simulation approach is being used to solve fleet sizing problem in flow shop production to ensure optimum situation. Optimum situation in this research means minimum flow time and maximum capacity in production floor. Simulation approach is being used because flow shop can be modelled into queuing network and inter-arrival time is not following exponential distribution. Therefore, contribution of this research is solving fleet sizing problem with multi objectives in flow shop production using simulation approach with ARENA Software

Fuel options for public bus fleets in Sweden

OpenAIRE

Xylia, Maria; Silveira, Semida

2015-01-01

The Swedish public transport sector has defined two major targets, i.e., to run 90% of the total vehicle kilometers of the fleet on non-fossil fuels and double the volume of travel via public transport by 2020, increasing the share of public transport in relation to the total personal transport in the country . The f3 report Fuel options for public bus fleets in Sweden highlights the challenges and solutions encountered, particularly when it comes to the adoption of renewable fuels in the reg...

Nuclear heat source component design considerations for HTGR process heat reactor plant concept

International Nuclear Information System (INIS)

McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

1982-01-01

Using alternate energy sources abundant in the U.S.A. to help curb foreign oil imports is vitally important from both national security and economic standpoints. Perhaps the most forwardlooking opportunity to realize national energy goals involves the integrated use of two energy sources that have an established technology base in the U.S.A., namely nuclear energy and coal. The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc.) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

Vehicular fleet operation on natural gas and propane: An overview. Final research report

International Nuclear Information System (INIS)

Taylor, D.B.; Mahmassani, H.; Euritt, M.A.

1992-11-01

The report attempts to contribute to the timely area of alternative vehicular fuels. It addresses the analysis of fleet operation on alternative fuels, specifically compressed natural gas (CNG) and propane, in terms of both fleet economics and societal impacts. Comprehensive information on engine technology, fueling infrastructure design, and societal impacts are presented. An evaluation framework useful for decisions between any vehicular fuels is developed. The comprehensive fleet cost-effectiveness analysis framework used in previous Project 983 reports is discussed in great detail. This framework/model is flexible enough to allow substantial sensitivity and scenario analysis. The model is used to perform sample analyses of both fleet economic and societal impacts

The Fleet Support Community: Meeting Its Mission in the 21st Century

National Research Council Canada - National Science Library

Murdy, Deanna

1999-01-01

This thesis evaluates the effectiveness of the Fleet Support community's management practices in meeting the dynamic changes in the complex fleet support arena, while increasing its value to the Navy in the future...

Versatile three-component procedure for combinatorial synthesis of ...

Indian Academy of Sciences (India)

component condensation reactions have been reported. The conventional synthesis involves a three-component condensation of isatin (or aromatic aldehyde) and malononitrile with dimedone or barbituric acid or 4- ... the capillary tube method with an electro thermal. 9200 apparatus. 1H NMR and 13C NMR spectra were.

BioFleet case studies

International Nuclear Information System (INIS)

2007-01-01

These six case studies examined the use of different biodiesel blends as fuel supply sources for businesses in British Columbia (BC). In the first case study, 6 municipalities participated in a pilot program designed to compare the performance of biodiesel and diesel fuels. Each municipality operated 2 base vehicles running on conventional diesel along with 2 similar vehicles which used biodiesel. Real time emissions tests and analyses of the vehicles using biodiesel were also conducted by 2 of the participating municipalities. All municipalities participating in the study agreed to purchase significant volumes of biodiesel. The second case study described a pilot study conducted by the City of Vancouver's equipment services branch in 2004. As a result of the study, the city now has over 530 types of equipment that use biodiesel. The third case study described a program designed by TSI Terminals in Vancouver to assess the emission reduction impact of using biodiesel at its port facility. Six different pieces of equipment were used to confirm that biodiesel could be used throughout the terminal. Test results confirmed that biodiesel blends could be used to reduce emissions. Overall emissions were reduced by 30 per cent. The fourth case study described a waste renderer that used a fleet of 36 trucks to deliver raw products to its plants. The company made the decision to use only biodiesel for its entire fleet of trucks. Since July 2005, the company has logged over 1.7 million km using biodiesel blends. The fifth case study described a salmon hatchery that switched from diesel to biodiesel in order to reduce emissions. The biodiesel blends are used to fuel the hatchery's 2 diesel generators. The hatchery has reduced emissions of greenhouse gases (GHGs) by an estimated 1800 tonnes annually. The sixth case study described how the Township of Langley has started using biodiesel for its entire fleet of of approximately 250 pieces of equipment. The township has not

Reactor pressure vessel and reactor coolant circuit cast duplex stainless steel components contribution of the expertise for life management studies

International Nuclear Information System (INIS)

Bezdikian, Georges

2006-09-01

The life management of French Nuclear Power Plants is a major stake from an economic and a technical point of view considering the aging management assessment of the key components of the plant. The actual life evaluation is the result of prediction of life assessment from important program of expertise for the 3-loop PWR and 4-loop PWR plants in operation. To optimize the strategic policy in order to achieve the best possible performance and to prepare the technical and economical choice and decision, the paper presents the association of life management strategy and the program of expertise considering: - the identification of degradation for different components and prediction criteria proposed; - the large database from cast reactor coolant and component removed from nuclear power plants and expertise studies to confirm the prediction; - the life evaluation of RPV with radiation surveillance program based on the expertise of irradiation capsules, it is particularly shown how the expertise is in the center of the strategic choice. The French utility has organized the life management of nuclear plant as a function of several programs of expertise of knowledge on the long term experience feedback and the maintenance program for life. This paper shows updated on RPV and reactor coolant equipment activities engaged by utility on: - periodic maintenance and volume of expertise; - Alternative maintenance actions; - Large volume of expertise and how are managed these results to predict the aging management. (author)

Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

International Nuclear Information System (INIS)

Poirier, J.; Dupouy, J.M.

Radiation effects on metals or alloys used in fast reactor core components are examined in the papers presented at this conference, the accent being put on swelling and irradiation creep of steels and nickel alloys

78 FR 64499 - Federal Acquisition Regulation; Submission for OMB Review; Contractor Use of Interagency Fleet...

Science.gov (United States)

2013-10-29

...; Submission for OMB Review; Contractor Use of Interagency Fleet Management System Vehicles AGENCY: Department... previously approved information collection requirement concerning contractor use of interagency fleet... comments identified by Information Collection 9000- 0032, Contractor Use of Interagency Fleet Management...

Vehicle attributes constraining present electric car applicability in the fleet market

Energy Technology Data Exchange (ETDEWEB)

Wagner, J R

1979-12-01

One strategy for reducing petroleum imports is to use electric cars in place of conventional vehicles. This paper examines obstacles which electric cars are likely to encounter in attempting to penetrate a key segment of the passenger car market, namely, the fleet market. A fleet is here defined as a group of cars operated by a corporation or a government agency. The primary data source is a questionnaire that was distributed to fleet operators by the Bobit Publishing Company in the summer of 1977. Six sectors of the fleet market were sampled: police, state and local government, utilities, taxi, rental, and business. The questionnaire was specifically designed to uncover factors limiting market penetration of unconventional vehicles, although no attempt was made to determine price elasticities. Emphasis is on vehicle attributes that are readily quantifiable and relatively projectable, including seating capacity, range, battery recharging characteristics, availability of power options, and ability to use interstate highways.

Fleet Aviation Maintenance Organic Support (FAMOS) Laboratory

Data.gov (United States)

Federal Laboratory Consortium — Purpose:The Fleet Aviation Maintenance Organic Support (FAMOS) Laboratory at the Naval Air Warfare Center Aircraft Division, Lakehurst, NJ provides rapid engineering...

Comparison of Simultaneous Nitrification and Denitrification for Three Different Reactors

Directory of Open Access Journals (Sweden)

W. Khanitchaidecha

2015-01-01

Full Text Available Discharge of high NH4-N containing wastewater into water bodies has become a critical and serious issue due to its negative impact on water and environmental quality. In this research, the performance of three different reactors was assessed and compared with regard to the removal of NH4-N from wastewater. The highest nitrogen removal efficiency of 98.3% was found when the entrapped sludge reactor (ESR, in which the sludge was entrapped in polyethylene glycol polymer, was used. Under intermittent aeration, nitrification and denitrification occurred simultaneously in the aerobic and anaerobic periods. Moreover, internal carbon was consumed efficiently for denitrification. On the other hand, internal carbon consumption was not found to occur in the suspended sludge reactor (SSR and the mixed sludge reactor (MSR and this resulted in nitrogen removal efficiencies of SSR and MSR being 64.7 and 45.1%, respectively. Nitrification and denitrification were the main nitrogen removal processes in the aerobic and anaerobic periods, respectively. However, due to the absence of sufficient organic carbon, denitrification was uncompleted resulting in high NO3-N contents in the effluent.

75 FR 20778 - Security Zone; Portland Rose Festival Fleet Week, Willamette River, Portland, OR

Science.gov (United States)

2010-04-21

...-AA87 Security Zone; Portland Rose Festival Fleet Week, Willamette River, Portland, OR AGENCY: Coast... during the Portland Rose Festival Fleet Week from June 2, 2010, through June 7, 2010. The security zone... is a need to provide a security zone for the 2010 Portland Rose Festival Fleet Week, and there is...

Civil ships propulsion reactor plants development and operation experience, and prospects for their improvement

International Nuclear Information System (INIS)

Vasyukov, V.I.; Kiryushin, A.I.; Panov, Yu.K.; Polunichev, V.I.

2000-01-01

Russia is alone country in the World possessing nuclear-powered icebreaker fleet. Phases of creation in Russia of several propulsion nuclear reactor plant generations are considered. By present 8 nuclear ice-breakers and a nuclear-powered cargo ship (lighter carrier) have been constructed, for which three of propulsion NSSS generations were developed. Their brief description, main performance indicators and results of operation since 1959 are given. It is shown that gradual evolution of NSSS design features has ensured creation of reliable, safe and environmentally friendly propulsion reactor plants. Issues of the propulsion NSSS life extension and improvement for new generation of nuclear ice-breakers, cargo ships, floating heat and power plants, sea water desalination and power generating complexes are considered with account for the gained operating experience. Activities on creation of new generation nuclear-powered ships and floating NPPs are considered as prospective sphere of Russia's collaboration with other countries of the World community. (author)

Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual. Part 3: Hardware component failure data; Volume 5, Revision 4

International Nuclear Information System (INIS)

Reece, W.J.; Gilbert, B.G.; Richards, R.E.

1994-09-01

This data manual contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 database, which is sponsored by the US Nuclear Regulatory Commission. NUCLARR was designed as a tool for risk analysis. Many of the nuclear reactors in the US and several outside the US are represented in the NUCLARR database. NUCLARR includes both human error probability estimates for workers at the plants and hardware failure data for nuclear reactor equipment. Aggregations of these data yield valuable reliability estimates for probabilistic risk assessments and human reliability analyses. The data manual is organized to permit manual searches of the information if the computerized version is not available. Originally, the manual was published in three parts. In this revision the introductory material located in the original Part 1 has been incorporated into the text of Parts 2 and 3. The user can now find introductory material either in the original Part 1, or in Parts 2 and 3 as revised. Part 2 contains the human error probability data, and Part 3, the hardware component reliability data

Executive Order 13514: Federal Leadership in Environmental, Energy, and Economic Performance; Comprehensive Federal Fleet Management Handbook (Book)

Energy Technology Data Exchange (ETDEWEB)

Daley, R. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Ahdieh, N. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Bentley, J. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

2014-01-01

A comprehensive Federal Fleet Management Handbook that builds upon the "Guidance for Federal Agencies on E.O. 13514 Section 12-Federal Fleet Management" and provides information to help fleet managers select optimal greenhouse gas and petroleum reduction strategies for each location, meeting or exceeding related fleet requirements, acquiring vehicles to support these strategies while minimizing fleet size and vehicle miles traveled, and refining strategies based on agency performance.

An Intelligent Fleet Condition-Based Maintenance Decision Making Method Based on Multi-Agent

Directory of Open Access Journals (Sweden)

Bo Sun

2012-01-01

Full Text Available According to the demand for condition-based maintenance online decision making among a mission oriented fleet, an intelligent maintenance decision making method based on Multi-agent and heuristic rules is proposed. The process of condition-based maintenance within an aircraft fleet (each containing one or more Line Replaceable Modules based on multiple maintenance thresholds is analyzed. Then the process is abstracted into a Multi-Agent Model, a 2-layer model structure containing host negotiation and independent negotiation is established, and the heuristic rules applied to global and local maintenance decision making is proposed. Based on Contract Net Protocol and the heuristic rules, the maintenance decision making algorithm is put forward. Finally, a fleet consisting of 10 aircrafts on a 3-wave continuous mission is illustrated to verify this method. Simulation results indicate that this method can improve the availability of the fleet, meet mission demands, rationalize the utilization of support resources and provide support for online maintenance decision making among a mission oriented fleet.

Tools for improving safety management in the Norwegian Fishing Fleet occupational accidents analysis period of 1998-2006.

Science.gov (United States)

Aasjord, Halvard L

2006-01-01

-July 2006 the numbers of fatal accidents and calculated risks are analysed for three main fleet groups. The highest risk factor of 24.8 fatal accidents per 10.000 man years is found in the smaller fleet, length of vessel (Loa) fleet (13 fleet (Loa > 28 meter).

The Retrofit Puzzle Extended: Optimal Fleet Owner Behavior over Multiple Time Periods

Science.gov (United States)

2009-08-04

In "The Retrofit Puzzle: Optimal Fleet Owner Behavior in the Context of Diesel Retrofit Incentive Programs" (1) an integer program was developed to model profit-maximizing diesel fleet owner behavior when selecting pollution reduction retrofits. Flee...

Anaerobic digestion of grain stillage at high organic loading rates in three different reactor systems

International Nuclear Information System (INIS)

Schmidt, Thomas; Pröter, Jürgen; Scholwin, Frank; Nelles, Michael

2013-01-01

In this study the anaerobic digestion of grain stillage in three different reactor systems (continuous stirred tank reactor, anaerobic sequencing batch reactor, fixed bed reactor) with and without immobilization of microorganisms was investigated to evaluate the performance during increase of the organic loading rate (OLR) from 1 to 10 g of volatile solids (VS) per liter reactor volume and day and decrease of the hydraulic retention time (HRT) from 40 to 6 days. No significant differences have been observed between the performances of the three examined reactor systems. The changes in OLR and HRT caused a reduction of the specific biogas production (SBP) of about 25% from about 650 to 550 L kg −1 of VS but would also diminish the necessary digester volume and investment costs of about 75% compared to the state of the art. -- Highlights: ► It was shown that without immobilization of microorganisms low HRT's are possible. ► No significant differences have been observed between different digester designs. ► Trace element supplementation is obligatory with grain stillage as substrate

Comparative energetics of three fusion-fission symbiotic nuclear reactor systems

International Nuclear Information System (INIS)

Gordon, C.W.; Harms, A.A.

1975-01-01

The energetics of three symbiotic fusion-fission nuclear reactor concepts are investigated. The fuel and power balances are considered for various values of systems parameters. The results from this analysis suggest that symbiotic fusion-fission systems are advantageous from the standpoint of economy and resource utilization. (Auth.)

Computer-controlled ultrasonic equipment for automatic inspection of nuclear reactor components after manufacturing

International Nuclear Information System (INIS)

Moeller, P.; Roehrich, H.

1983-01-01

After foundation of the working team ''Automated US-Manufacture Testing'' in 1976 the realization of an ultrasonic test facility for nuclear reactor components after manufacturing has been started. During a period of about 5 years, an automated prototype facility has been developed, fabricated and successfully tested. The function of this facility is to replace the manual ultrasonic tests, which are carried out autonomically at different stages of the manufacturing process and to fulfil the test specification under improved economic conditions. This prototype facility has been designed as to be transported to the components to be tested at low expenditure. Hereby the reproduceability of a test is entirely guaranteed. (orig.) [de

The Fleet-Management as an Element of the Modern Development in the Ukrainian Leasing Market

OpenAIRE

Garafonova Olga I.; Svynarchuk Anastasiia V.

2017-01-01

The article is aimed at researching practical aspects of the fleet-management development in Ukraine; relationships that arise in this process; elements that interact in the market. The preconditions of the companies’ transition to the use of this concept in the management of fleet, as well as the opportunities that arise in the enterprises using the principles of fleet-management were considered. The main advantages and disadvantages of the implementation of fleet-management in companies wit...

Using fleets of electric-drive vehicles for grid support

International Nuclear Information System (INIS)

Tomic, Jasna; Kempton, Willett

2007-01-01

Electric-drive vehicles can provide power to the electric grid when they are parked (vehicle-to-grid power). We evaluated the economic potential of two utility-owned fleets of battery-electric vehicles to provide power for a specific electricity market, regulation, in four US regional regulation services markets. The two battery-electric fleet cases are: (a) 100 Th.nk City vehicle and (b) 252 Toyota RAV4. Important variables are: (a) the market value of regulation services, (b) the power capacity (kW) of the electrical connections and wiring, and (c) the energy capacity (kWh) of the vehicle's battery. With a few exceptions when the annual market value of regulation was low, we find that vehicle-to-grid power for regulation services is profitable across all four markets analyzed. Assuming now more than current Level 2 charging infrastructure (6.6 kW) the annual net profit for the Th.nk City fleet is from US$ 7000 to 70,000 providing regulation down only. For the RAV4 fleet the annual net profit ranges from US$ 24,000 to 260,000 providing regulation down and up. Vehicle-to-grid power could provide a significant revenue stream that would improve the economics of grid-connected electric-drive vehicles and further encourage their adoption. It would also improve the stability of the electrical grid. (author)

FLEET Velocimetry Measurements on a Transonic Airfoil

Science.gov (United States)

Burns, Ross A.; Danehy, Paul M.

2017-01-01

Femtosecond laser electronic excitation tagging (FLEET) velocimetry was used to study the flowfield around a symmetric, transonic airfoil in the NASA Langley 0.3-m TCT facility. A nominal Mach number of 0.85 was investigated with a total pressure of 125 kPa and total temperature of 280 K. Two-components of velocity were measured along vertical profiles at different locations above, below, and aft of the airfoil at angles of attack of 0 deg, 3.5 deg, and 7deg. Measurements were assessed for their accuracy, precision, dynamic range, spatial resolution, and overall measurement uncertainty in the context of the applied flowfield. Measurement precisions as low as 1 m/s were observed, while overall uncertainties ranged from 4 to 5 percent. Velocity profiles within the wake showed sufficient accuracy, precision, and sensitivity to resolve both the mean and fluctuating velocities and general flow physics such as shear layer growth. Evidence of flow separation is found at high angles of attack.

Business fleet refueling assessment. Final report, March 1992-November 1992

International Nuclear Information System (INIS)

Chaudier, A.

1993-01-01

The report investigates refueling characteristics and capabilities for the light-duty market sector, particularly business automobile fleets. It develops a profile of the characteristics of light-duty and automobile business fleets in 22 designated non-attainment areas, as well as Pittsburgh and Salt Lake City. Reliable data on these subjects will enable better analysis of the market potential for natural gas in the surveyed areas

Final Report: Safety of Plasma Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

Energy Technology Data Exchange (ETDEWEB)

Bourham, Mohamed A.; Gilligan, John G.

1999-08-14

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.

Analysis of transition to fuel cycle system with continuous recycling in fast and thermal reactors - 5060

International Nuclear Information System (INIS)

Passereini, S.; Feng, B.; Fei, T.; Kim, T.K.; Taiwo, T.A.; Brown, N.R.; Cuadra, A.

2015-01-01

A recent Evaluation and Screening study of nuclear fuel cycle options identified a few groups of options as most promising. One of these most promising Evaluation Groups (EGs) is characterized by the continuous recycling of uranium (U) and transuranics (TRU) with natural uranium feed in both fast and thermal critical reactors. This evaluation group, designated as EG30, is represented by an example fuel cycle option that employs a two-technology, two-stage fuel cycle system. The first stage involves the continuous recycling of co-extracted U/TRU in Sodium-cooled Fast Reactors (SFRs) with metallic fuel and breeding ratio greater than 1. The second stage involves the use of the surplus TRU in Mixed Oxide (MOX) fuel in Pressurized Water Reactors that are MOX-capable (MOX-PWRs). This paper presents and discusses preliminary fuel cycle analysis results from the fuel cycle codes VISION and DYMOND for the transition to this fuel cycle option from the current once-through cycle in the United States (U.S.) that consists of Light Water Reactors (LWRs) that only use conventional UO 2 fuel. The analyses in this paper are applicable for a constant 100 GWe capacity, roughly the size of the U.S. nuclear fleet. Two main strategies for the transition to EG30 were analyzed: 1) deploying both SFRs and MOX-PWRs in parallel or 2) deploying them in series with the SFR fleet first. With an estimated retirement schedule for the existing LWRs, an assumed reactor lifetime of 60 years, and no growth, the nuclear system fully transitions to the new fuel cycle within 100 years for both strategies without SFR fuel shortages. Compared to the once-through cycle, transition to the SFR/MOX-PWR fleet with continuous recycle was shown to offer significant reductions in uranium consumption and waste disposal requirements. In addition, these initial calculations revealed a few notable modeling and strategy questions regarding how recycled resources are allocated, reactors that can switch between

Fleet Purchase Behavior: Decision Processes and Implications for New Vehicle Technologies and Fuels

OpenAIRE

Nesbitt, Kevin; Sperling, Daniel

2001-01-01

Vehicle fleets are a poorly understood part of the economy. They are important, though, in that they purchase a large share of light-duty vehicles and are often targeted by governments as agents of change. We investigate fleet purchase behavior, using focus groups, interviews, and mail and telephone surveys. We categorize fleets into four different decision-making structures (autocratic, bureaucratic, hierarchic, and democratic), determine what share of the market sector each represents, d...

Coatings for fast breeder reactor components

International Nuclear Information System (INIS)

Johnson, R.N.

1984-04-01

Several types of metallurgical coatings are used in the unique environments of the fast breeder reactor. Most of the coatings have been developed for tribological applications, but some also serve as corrosion barriers, diffusion barriers, or radionuclide traps. The materials that have consistently given the best performance as tribological coatings in the breeder reactor environments have been coatings based on chromium carbide, nickel aluminide, or Tribaloy 700 (a nickel-base hard-facing alloy). Other coatings that have been qualified for limited applications include chromium plating for low temperature galling protection and nickel plating for radionuclide trapping

Nonlinear dynamic analysis of nuclear reactor primary coolant systems

International Nuclear Information System (INIS)

Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.

1979-01-01

The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development

Reactor System Design

International Nuclear Information System (INIS)

Chi, S. K.; Kim, G. K.; Yeo, J. W.

2006-08-01

SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

Multi-Year On-Road Emission Factor Trends of Two Heavy-Duty California Fleets

Science.gov (United States)

Haugen, M.; Bishop, G.

2017-12-01

New heavy-duty vehicle emission regulations have resulted in the development of advanced exhaust after-treatment systems that specifically target particulate matter (PM) and nitrogen oxides (NOx = NO + NO2). This has resulted in significant decreases in the emissions of these species. The University of Denver has collected three data sets of on-road gaseous (CO, HC, NO and NOx) and PM (particle mass, black carbon and particle number) emission measurements from heavy-duty vehicles (HDVs) in the spring of 2013, 2015 and 2017 at two different locations in California. One site is located at the Port of Los Angeles, CA (1,150 HDVs measured in 2017) and the other site is located at a weigh station in Northern California near Cottonwood, CA (780 HDVs measured in 2017). The On-Road Heavy-Duty Measurement Setup measures individual HDV's fuel specific emissions (DOI: 10.1021/acs.est.6b06172). Vehicles drive under a tent-like structure that encapsulates vehicle exhaust and 15 seconds of data collection is integrated to give fuel specific information. The measurements obtained from these campaigns contain real-world emissions affected by different driving modes, after-treatment systems and location. The Port of Los Angeles contributes a fleet that is fully equipped with diesel particulate filters (DPFs) as a result of the San Pedro Ports Clean Air Action Plan enforced since 2010 that allows only vehicles model year 2007 or newer on the premises. This fleet, although comprised with relatively new HDVs with lower PM emissions, has increased PM emissions as it has aged. Cottonwood's fleet contains vehicles with and without after-treatment systems, a result of a gradual turnover rate, and fleet PM has decreased at a slower rate than at the Port of Los Angeles. The decrease in PM emissions is a result of more HDVs being newer model years as well as older model years being retrofit with DPFs. The complimentary fleets, studied over multiple years, have given the University of Denver

Product Characterization and Kinetics of Biomass Pyrolysis in a Three-Zone Free-Fall Reactor

Directory of Open Access Journals (Sweden)

Natthaya Punsuwan

2014-01-01

Full Text Available Pyrolysis of biomass including palm shell, palm kernel, and cassava pulp residue was studied in a laboratory free-fall reactor with three separated hot zones. The effects of pyrolysis temperature (250–1050°C and particle size (0.18–1.55 mm on the distribution and properties of pyrolysis products were investigated. A higher pyrolysis temperature and smaller particle size increased the gas yield but decreased the char yield. Cassava pulp residue gave more volatiles and less char than those of palm kernel and palm shell. The derived solid product (char gave a high calorific value of 29.87 MJ/kg and a reasonably high BET surface area of 200 m2/g. The biooil from palm shell is less attractive to use as a direct fuel, due to its high water contents, low calorific value, and high acidity. On gas composition, carbon monoxide was the dominant component in the gas product. A pyrolysis model for biomass pyrolysis in the free-fall reactor was developed, based on solving the proposed two-parallel reactions kinetic model and equations of particle motion, which gave excellent prediction of char yields for all biomass precursors under all pyrolysis conditions studied.

Fleet Assistance and Support Team (FAST) Lab

Data.gov (United States)

Federal Laboratory Consortium — The FAST team was established by PMA-264 for introduction of multistatic ASW systems into the Fleet.FAST provides Air ASW mission planning, tactics/tactical sensor...

Formation of optimal construction fleet composition

Science.gov (United States)

Tuskaeva, Zalina

2017-10-01

Machinery supply and its rational use in construction processes considerably determine the final product of construction organizations. Therefore, the problem of defining the type size composition of the construction fleet as one of the lowest material-intensive productions, is of a particular importance.

75 FR 8563 - Safety Zone; Fleet Week Maritime Festival, Pier 66, Elliott Bay, Seattle, WA

Science.gov (United States)

2010-02-25

...-AA00 Safety Zone; Fleet Week Maritime Festival, Pier 66, Elliott Bay, Seattle, WA AGENCY: Coast Guard... Fleet Week Maritime Festival. This safety zone is necessary as these events have historically resulted... the safety of life and property on navigable waters during the annual Fleet Week Maritime Festival...

Advanced technology mobile robotics vehicle fleet

International Nuclear Information System (INIS)

McGovern, D.E.

1987-03-01

A fleet of vehicles is being developed and maintained by Sandia National Laboratories for studies in remote control and autonomous operation. The vehicles range from modified commercial vehicles to specially constructed mobile platforms and are utilized as testbeds for developing concepts in the areas of remote control (teleoperation) and computer control (autonomy). Actuators control the vehicle speed, brakes, and steering via manual input from a remote driving station or through some level of digital computer control. On-board processing may include simple vehicle control functions or may allow for unmanned, autonomous operation. Communication links are provided for digital communication between control computers, television transmission for vehicle vision, and voice for local control. SNL can develop, test, and evaluate sensors, processing requirements, various methods of actuator implementation, operator controlled feedback requirements, and vehicle operations. A description of the major features and uses for each of the vehicles in the fleet is provided

Optimal Routing for Heterogeneous Fixed Fleets of Multicompartment Vehicles

Directory of Open Access Journals (Sweden)

Qian Wang

2014-01-01

Full Text Available We present a metaheuristic called the reactive guided tabu search (RGTS to solve the heterogeneous fleet multicompartment vehicle routing problem (MCVRP, where a single vehicle is required for cotransporting multiple customer orders. MCVRP is commonly found in delivery of fashion apparel, petroleum distribution, food distribution, and waste collection. In searching the optimum solution of MCVRP, we need to handle a large amount of local optima in the solution spaces. To overcome this problem, we design three guiding mechanisms in which the search history is used to guide the search. The three mechanisms are experimentally demonstrated to be more efficient than the ones which only apply the known distance information. Armed with the guiding mechanisms and the well-known reactive mechanism, the RGTS can produce remarkable solutions in a reasonable computation time.

Assessing the Future Vehicle Fleet Electrification: The Impacts on Regional and Urban Air Quality.

Science.gov (United States)

Ke, Wenwei; Zhang, Shaojun; Wu, Ye; Zhao, Bin; Wang, Shuxiao; Hao, Jiming

2017-01-17

There have been significant advancements in electric vehicles (EVs) in recent years. However, the different changing patterns in emissions at upstream and on-road stages and complex atmospheric chemistry of pollutants lead to uncertainty in the air quality benefits from fleet electrification. This study considers the Yangtze River Delta (YRD) region in China to investigate whether EVs can improve future air quality. The Community Multiscale Air Quality model enhanced by the two-dimensional volatility basis set module is applied to simulate the temporally, spatially, and chemically resolved changes in PM 2.5 concentrations and the changes of other pollutants from fleet electrification. A probable scenario (Scenario EV1) with 20% of private light-duty passenger vehicles and 80% of commercial passenger vehicles (e.g., taxis and buses) electrified can reduce average PM 2.5 concentrations by 0.4 to 1.1 μg m -3 during four representative months for all urban areas of YRD in 2030. The seasonal distinctions of the air quality impacts with respect to concentration reductions in key aerosol components are also identified. For example, the PM 2.5 reduction in January is mainly attributed to the nitrate reduction, whereas the secondary organic aerosol reduction is another essential contributor in August. EVs can also effectively assist in mitigating NO 2 concentrations, which would gain greater reductions for traffic-dense urban areas (e.g., Shanghai). This paper reveals that the fleet electrification in the YRD region could generally play a positive role in improving regional and urban air quality.

Development and application of computer codes for multidimensional thermalhydraulic analyses of nuclear reactor components

International Nuclear Information System (INIS)

Carver, M.B.

1983-01-01

Components of reactor systems and related equipment are identified in which multidimensional computational thermal hydraulics can be used to advantage to assess and improve design. Models of single- and two-phase flow are reviewed, and the governing equations for multidimensional analysis are discussed. Suitable computational algorithms are introduced, and sample results from the application of particular multidimensional computer codes are given

Meso-scale magnetic signatures for nuclear reactor steel irradiation embrittlement monitoring

Energy Technology Data Exchange (ETDEWEB)

Suter, J. D., E-mail: pradeep.ramuhalli@pnnl.gov; Ramuhalli, P., E-mail: pradeep.ramuhalli@pnnl.gov; Hu, S.; Li, Y.; Jiang, W.; Edwards, D. J.; Schemer-Kohrn, A. L.; Johnson, B. R. [Pacific Northwest National Laboratory, 902 Battelle Blvd, Richland, WA 99352 (United States); McCloy, J. S., E-mail: john.mccloy@wsu.edu; Xu, K., E-mail: john.mccloy@wsu.edu [Washington State University, PO Box 642920, Pullman, WA 99164 (United States)

2015-03-31

Verifying the structural integrity of passive components in light water and advanced reactors will be necessary to ensure safe, long-term operations of the existing U.S. nuclear fleet. This objective can be achieved through nondestructive condition monitoring techniques, which can be integrated with plant operations to quantify the “state of health” of structural materials in real-time. While nondestructive methods for monitoring many classes of degradation (such as fatigue or stress corrosion cracking) are relatively advanced, this is not the case for degradation caused by irradiation. The development of nondestructive evaluation technologies for these types of degradation will require advanced materials characterization techniques and tools that enable comprehensive understanding of nuclear reactor material microstructural and behavioral changes under extreme operating environments. Irradiation-induced degradation of reactor steels causes changes in their microstructure that impacts their micro-magnetic properties. In this paper, we describe preliminary results of integrating advanced material characterization techniques with meso-scale computational models. In the future, this will help to provide an interpretive understanding of the state of degradation in structural materials. Microstructural data are presented from monocrystalline Fe and are correlated with variable-field magnetic force microscopy and micro-magnetic measurements. Ongoing research is focused on extending the measurements and models on thin films to gain insights into the structural state of irradiated materials and the resulting impact on magnetic properties. Preliminary conclusions from these correlations are presented, and next steps described.

RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1986

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1986-01-01

In order to enable future reliable operation of the RA reactor, according to new licensing regulations, three major tasks started in 1984 were fulfilled: building of the new emergency system, reconstruction of the existing ventilation system, and reconstruction of the power supply system. Simultaneously in 1985/1986 renewal of the instrumentation and reconstruction of the system for handling and storage of the spent fuel in the reactor building have started. Design projects for these tasks are almost finished and the reconstruction of both systems is expected to be finished until 1988 and mid 1989 respectively. RA reactor Safety report was finished according to the recommendations of the IAEA. Investments in 1986 were used for 8000 kg of heavy water, maintenance of reactor systems and supply of new components, reconstruction of reactor systems. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr