Energy Research Advisory Board, Civilian Nuclear Power Panel: Subpanel 1 report, Light water reactor utilization and improvement: Volume 2

International Nuclear Information System (INIS)

1986-10-01

The Secretary of Energy requested that the Office of Nuclear Energy prepare a strategic national plan that outlines the Department's role in the future development of civilian nuclear power and that the Energy Research Advisory Board establish an ad hoc panel to review and comment on this plan. The Energy Research Advisory Board formed a panel for this review and three subpanels were formed. One subpanel was formed to address the institutional issues surrounding nuclear power, one on research and development for advanced nuclear power plants and a third subpanel on light water reactor utilization and improvement. The subpanel on light water reactors held two meetings at which representatives of the DOE, the NRC, EPRI, industry and academic groups made presentations. This is the report of the subpanel on light water reactor utilization and improvement. This report presents the subpanel's assessment of initiatives which the Department of Energy should undertake in the national interest, to develop and support light water reactor technologies

Standardization of advanced light water reactors and progress on achieving utility requirements

International Nuclear Information System (INIS)

Marston, T.U.; Layman, W.H.; Bockhold, G. Jr.

1992-01-01

This paper reports that for a number of years, the U.S. utilities had led an industry-wide effort to establish a technical foundation for the design of the next generation of light water reactors in the United States. Since 1985, this utility initiative has been effected through a major technical program managed by the Electric Power Research Institute (EPRI); the U.S. Advanced Light Water Reactor (ALWR) Program. In addition to the U.S. utility leadership and sponsorship, the ALWR Program also has the participation and sponsorship of a number of international utility companies and close cooperation with the U.S. Department of Energy (DOE). The NPOC Strategic Plan for Building New Nuclear Plants creates a framework within which new standardized nuclear plants may be built. The Strategic Plan is an expression of the nuclear energy industry's serious intent to create the necessary conditions for new plant construction and operation. The industry has assembled a comprehensive, integrated list of actions that must be taken before new plants will be built and assigns responsibility for managing the various issues and sets time-tables and milestones against which we must measure progress

Safety of light water reactors. Risks of nuclear technology

International Nuclear Information System (INIS)

Veser, Anke; Schlueter, Franz-Hermann; Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen; Kessler, Guenter

2012-01-01

The book on the safety of light-water reactors includes the following chapters: Part I: Physical and technical safety concept of actual German and future European light-water reactors: (1) Worldwide operated nuclear power plants in 2011, (2) Some reactor physical fundamentals. (3) Nuclear power plants in Germany. (4) Radioactive exposure due to nuclear power plants. (5) Safety concept of light-water reactors. (6) Probabilistic analyses and risk studies. (7) Design of light-water reactors against external incidents. (8) Risk comparison of nuclear power plants and other energy systems. (9) Evaluation of risk studies using the improved (new) safety concept for LWR. (19) The severe reactor accidents of Three Mile Island, Chernobyl and Fukushima. Part II: Safety of German LWR in case of a postulated aircraft impact. (11) Literature. (12) Review of requirements and actual design. (13) Incident scenarios. (14) Load approach for aircraft impact. (15) Demonstration of the structural behavior in case of aircraft impact. (16) Special considerations. (17) Evaluation of the safety state of German and foreign nuclear power plants. Part III: ROSOS as example for a computer-based decision making support system for the severe accident management. (19) Literature. (20) Radiological fundamentals, accident management, modeling of the radiological situation. (21) The decision making support system RODOS. (22) RODOS and the Fukushima accident. (23) Recent developments in the radiological emergency management in the European frame.

Utility Leadership in Defining Requirements for Advanced Light Water Reactors

International Nuclear Information System (INIS)

Sugnet, William R.; Layman, William H.

1990-01-01

It is appropriate, based on twenty five years of operating experience, that utilities take a position of leadership in developing the technical design and performance requirements for the next generations of nuclear electric generating plants. The U. S. utilities, through the Electric Power Research Institute, began an initiative in 1985 to develop such Utility requirements. Many international Utility organizations, including Korea Electric Power Corporation, have joined as full participants in this important Utility industry initiative. In light of the closer linkage among countries of the world due to rapid travel and telecommunications, it is also appropriate that there be international dialogue and agreement on the principal standards for nuclear power plant acceptability and performance. The Utility/EPRI Advanced Light Water Reactor Program guided by the ALRR Utility Steering Committee has been very successful in developing these Utility requirements. This paper will summarize the state of development of the ALRR Utility Requirements for Evolutionary Plants, recent developments in their review by the U. S. Nuclear Regulatory Commission, resolution of open issues, and the extension of this effort to develop a companion set of ALRR Utility Requirements for plants employing passive safety features

NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapter 1, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapters 2--13, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)

Overview of light water/hydrogen-based low energy nuclear reactions

International Nuclear Information System (INIS)

Miley, George H.; Shrestha, Prajakti J.

2006-01-01

This paper reviews light water and hydrogen-based low-energy nuclear reactions (LENRs) including the different methodologies used to study these reactions and the results obtained. Reports of excess heat production, transmutation reactions, and nuclear radiation emission are cited. An aim of this review is to present a summary of the present status of light water LENR research and provide some insight into where this research is heading. (author)

Advanced light water reactors: an economically viable part of the world's future energy mix

International Nuclear Information System (INIS)

Bruschi, H.J.

1996-01-01

In addition to safety and reliability, a common mission for the international nuclear industry in the 21. century will be ensure affordable electricity. At the Westinghouse Electric Corporation believe our advanced light water reactor (ALWR) design gives us the opportunity to provide the safest, most reliable, lowest cost, most competitive generation method possible for use by nations and utilities worldwide. While the safety and reliability aspects of the ALWR can be proven tangibly and are well-documented, questions have been raised about the technology's ability to work within the world's selling price range for electricity generation. For our industry's financial stability, and especially for the stability of the world's future power needs, Westinghouse has done extensive work on this issue and we are convinced we can meet the competitive challenge. We believe the ALWR can be an economically viable part of the world's future energy mix. This paper will define the competitive challenge that is being addressed by the industry and then analyze three specific areas: capital costs, operating costs, and financing costs. The hidden advantage of nuclear power in responding to these challenges will be explored, and a strong case will be made asserting that the advanced light water reactor will be able to compete in international markets with viable production costs. (authors)

Passive autocatalytic recombiners for combustible gas control in advanced light water reactors

International Nuclear Information System (INIS)

Wolff, U.; Sliter, G.

2004-01-01

A key aspect of the worldwide effort to develop advanced nuclear power plants is designing to address severe accident phenomena, including the generation of hydrogen during core melt progression (metal-water and core-concrete reactions). This design work not only resolves safety concerns with hydrogen, but also supports the development of a technical basis for simplification of off-site emergency planning. The dominant challenge to any emergency planning approach is a large, early containment failure due to pressure excursions. Among the potential contributors to large and rapid increases in containment pressure is hydrogen combustion. The more improbable a containment-threatening combustion becomes, the more appropriate the argument for significant emergency planning simplification. As discussed in this paper, catalytic recombiners provide a means to passively and reliably limit hydrogen combustion to a continuous oxidation process with virtually no potential for containment failure in passive advanced light water reactors (ALWRs). (author)

Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

Energy Technology Data Exchange (ETDEWEB)

Professor Neill Todreas

2001-10-01

A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team

Reliability assurance programme guidebook for advanced light water reactors

International Nuclear Information System (INIS)

2001-12-01

To facilitate the implementation of reliability assurance programmes (RAP) within future advanced reactor programmes and to ensure that the next generation of commercial nuclear reactors achieves the very high levels of safety, reliability and economy which are expected of them, in 1996, the International Atomic Energy Agency (IAEA) established a task to develop a guidebook for reliability assurance programmes. The draft RAP guidebook was prepared by an expert consultant and was reviewed/modified at an Advisory Group meeting (7-10 April 1997) and at a consults meeting (7-10 October 1997). The programme for the RAP guidebook was reported to and guided by the Technical Working Group on Advanced Technologies for Light Water Reactors (TWG-LWR). This guidebook will demonstrate how the designers and operators of future commercial nuclear plants can exploit the risk, reliability and availability engineering methods and techniques developed over the past two decades to augment existing design and operational nuclear plant decision-making capabilities. This guidebook is intended to provide the necessary understanding, insights and examples of RAP management systems and processes from which a future user can derive his own plant specific reliability assurance programmes. The RAP guidebook is intended to augment, not replace, specific reliability assurance requirements defined by the utility requirements documents and by individual nuclear steam supply system (NSSS) designers. This guidebook draws from utility experience gained during implementation of reliability and availability improvement and risk based management programmes to provide both written and diagrammatic 'how to' guidance which can be followed to assure conformance with the specific requirements outlined by utility requirements documents and in the development of a practical and effective plant specific RAP in any IAEA Member State

Advanced light water reactors for the nineties

International Nuclear Information System (INIS)

Ross, F.A.; Sugnet, W.R.

1987-01-01

The EPRI/Industry advanced light water reactor (ALWR) program and the US Department of Energy ALWR program are closely coordinated to meet the common objective which is the availability of improved and simplified light water reactor plants that may be ordered in the next decade to meet new or replacement capacity requirements. The EPRI/Industry objectives, program participants, and foreign participants, utility requirements document, its organization and content, small plant conceptual design program, the DOE ALWR program, design verification program, General Electric ABWR design features, Combustion Engineering system design, mid-size plant development, General Electric SBWR objectives, Westinghouse/Burns and Roe design objectives, construction improvement, and improved instrumentation and control are discussed in the paper

GE's advanced nuclear reactor designs

International Nuclear Information System (INIS)

Berglund, R.C.

1993-01-01

The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

A collection of publications and articles for a light water ultra-safe plant concept

International Nuclear Information System (INIS)

Klevans, E.H.

1988-01-01

This collection contains reports titled: ''The Penn State Ultra-Safe Reactor Concept; '' ''Ultra Safe Nuclear Power; '' ''Use of the Modular Modeling System, in the Design of the Penn State Advanced Light Water Reactor; '' ''Use of the Modular Modeling System in Severe Transient Analysis of Penn State Advanced Light Water Reactor; '' ''PSU Engineers' Reactor Design May Stop a Future TMI; '' and ''The Penn State Advanced Light Water reactor Concept.''

Needs of nuclear data for advanced light water reactor

International Nuclear Information System (INIS)

Chaki, Masao

2008-01-01

Hitachi has been developing medium sized ABWRs as a power source that features flexibility to meet various market needs, such as minimizing capital risks, providing a timely return on capital investments, etc. Basic design concepts of the medium sized ABWRs are 1) using the current ABWR design which has accumulated favorable construction and operation histories as a starting point; 2) utilizing standard BWR fuels which have been fabricated by proven technology; 3) achieving a rationalized design by suitably utilizing key components developed for large sized reactors. Development of the medium sized ABWRs has proceeded in a systematic, stepwise manner. The first step was to design an output scale for the 600MWe class reactor (ABWR-600), and the next step was to develop an uprating concept to extend this output scale to the 900MWe class reactor (ABWR-900) based on the rationalized technology of the ABWR-600 for further cost savings. In addition, Hitachi and MHI developed an ultra small reactor, 'Package-Reactor'. About the nuclear data, for the purpose of verification of the nuclear analysis method of BWR for mixed oxide (MOX) cores, UO 2 and MOX fuel critical experiments EPICURE and MISTRAL were analyzed using nuclear design codes HINES and CERES with ENDF/B nuclear data file. The critical keffs of the absorber worth experiments, the water hole worth experiments and the 2D void worth experiments agreed with those of the reference experiments within about 0.1%Δk. The root mean square differences of radial power distributions between calculation and measurement were almost less than 2.0%. The calculated reactivity worth values of the absorbers, the water hole and the 2D void agreed with the measured values within nearly experimental uncertainties. These results indicate that the nuclear analysis method of BWR in the present paper give the same accuracy for the UO 2 cores and the MOX cores. (author)

Factors in the economic viability of advanced light water reactors

International Nuclear Information System (INIS)

Matzie, R.A.; Bagnal, C.W.; Rohde, K.R.

1997-01-01

Nuclear power currently produces over 20% of the electricity generated in the United States, and a similar number for the entire world. Electricity generated from these nuclear power plants is typically some of the most economical of all sources, and is becoming even more economical with time as utilities focus on reducing production costs. Nevertheless, with the exception of the Asia Pacific region, no new nuclear orders have been placed in many years, and none are planned for the forseeable future. Two reasons for this demise for nuclear power in the western world are usually put forward: the current price of alternative means of electric power generation and the political climate, which tends to be anti-nuclear. The first of these reasons is founded in the low price of natural gas, which has been the preferred fuel for recent power generation additions. These additions have principally been used as peaking units, which are required only at the highest demand periods and not as base load units. The second reason stems from some bad experiences in the post-TMI era, when projects experienced a rapidly changing regulatory environment, long schedule stretchouts, and huge cost overruns. In spite of this relatively poor environment for new nuclear power plants, major programs to develop advanced light water reactors are continuing to keep the nuclear option alive, both in the United States and Europe. These programs are aimed at capturing the lessons learned from past experience, to ensure the success of future nuclear projects. 6 refs., 8 figs., 1 tab

Plan for research to improve the safety of light-water nuclear power plants

International Nuclear Information System (INIS)

1978-03-01

This is the U.S. Nuclear Regulatory Commission's first annual report to Congress on recommendations for research on improving the safety of light-water nuclear power plants. Suggestions for reactor safety research were identified in, or received from, various sources, including the Advisory Committee on Reactor Safeguards, the NRC regulatory staff, and the consultants to the Research Review Group. After an initial screening to eliminate those not related to improved reactor safety, all the suggestions were consolidated into research topics. It is recommended that the following research projects be carried out: alternate containment concepts, especially vented containments; alternate decay heat removal concepts, especially add-on bunkered systems; alternate emergency core cooling concepts; improved in-plant accident response; and advanced seismic designs

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

Nuclear safety in light water reactors severe accident phenomenology

CERN Document Server

Sehgal, Bal Raj

2011-01-01

This vital reference is the only one-stop resource on how to assess, prevent, and manage severe nuclear accidents in the light water reactors (LWRs) that pose the most risk to the public. LWRs are the predominant nuclear reactor in use around the world today, and they will continue to be the most frequently utilized in the near future. Therefore, accurate determination of the safety issues associated with such reactors is central to a consideration of the risks and benefits of nuclear power. This book emphasizes the prevention and management of severe accidents to teach nuclear professionals

Economic benefits of advanced materials in nuclear power systems

International Nuclear Information System (INIS)

Busby, J.T.

2009-01-01

A key obstacle to the commercial deployment of advanced fast reactors is the capital cost. There is a perception of higher capital cost for fast reactor systems than advanced light water reactors. However, cost estimates come with a large uncertainty since far fewer fast reactors have been built than light water reactor facilities. Furthermore, the large variability of industrial cost estimates complicates accurate comparisons. Reductions in capital cost can result from design simplifications, new technologies that allow reduced capital costs, and simulation techniques that help optimize system design. It is plausible that improved materials will provide opportunities for both simplified design and reduced capital cost. Advanced materials may also allow improved safety and longer component lifetimes. This work examines the potential impact of advanced materials on the capital investment cost of fast nuclear reactors.

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Science.gov (United States)

2010-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

Nuclear Data Libraries for Hydrogen in Light Water Ice

International Nuclear Information System (INIS)

Torres, L; Gillette, V.H

2000-01-01

Nuclear data libraries were produced for hydrogen (H) in light water ice at different temperatures, 20, 30, 50, 77, 112, 180, 230 K.These libraries were produced using the NJOY nuclear data processing system.With this code we produce pointwise cross sections and related quantities, in the ENDF format, and in the ACE format for MCNP.Experimental neutron spectra at such temperatures were compared with MCNP4B simulations, based on the locally produced libraries, leading to satisfactory results

Indian Nuclear Society annual conference-1994 on advanced technologies related to nuclear power: proceedings

International Nuclear Information System (INIS)

Grover, R.B.

1994-01-01

The focal theme of the conference is advanced technologies related to nuclear power. Over the past three decades civilian nuclear power plants around the world have accumulated about 6000 reactor years of experience and have performed quite well. Overall safety record has been satisfactory. However, nuclear community is trying to compete with its own record by trying to enhance the safety characteristics of the best operating plant. A safety culture has been established in the nuclear establishments, which is providing impetus to advances in all aspects of nuclear technology all over the world. India has ongoing programmes for the development of advanced reactors and related advanced technologies. Evolution of pressurised heavy water reactors in India, developments made in the design of advanced heavy water reactor and the fast reactor programme, are some of the topics covered in addition to highlighting worldwide developments for the next generation of light water reactors. India is one of the few countries in the world where expertise about complete fuel cycle is available. Developments in the back end of the fuel cycle, use of thorium and plutonium and other related issues are also discussed. Technology control regimes being advocated and adopted by developed nations make it imperative for us to indigenise every equipment and component that goes into a power plant. In view of this, some aspects of manufacturing technologies, inspection techniques and maintenance problems are also covered. Relevant papers are processed separately for INIS. (M.K.V.)

Coatings used in light-water nuclear power plants

International Nuclear Information System (INIS)

Anon.

1981-01-01

The guide is intended to provide a common basis in the selection of test methods which may be required to evaluate and qualify protective coatings (paints) to be used in a light-water nuclear power plant. Standard test methods for the determination of fire resistance, chemical resistance, physical properties, effects of radiation, decontaminability, thermal conductivity, repairability, and for evaluation under accident conditions are included

Maintenance Cycle Extension in the IRIS Advanced Light Water Reactor Plant Design

International Nuclear Information System (INIS)

Galvin, Mark R.; Todreas, Neil E.; Conway, Larry E.

2003-01-01

New nuclear power generation in the United States will be realized only if the economic performance can be made competitive with other methods of electrical power generation. The economic performance of a nuclear power plant can be significantly improved by increasing the time spent on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described that can be used to resolve, in the design phase, maintenance-related operating cycle length barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the International Reactor, Innovative and Secure (IRIS) design. IRIS is an advanced light water nuclear power plant that is being designed to maximize this on-line generating time by increasing the operating cycle length. This is consequently a maintenance strategy paper using the IRIS plant as the example.Potential IRIS operating cycle length maintenance-related barriers, determined by modification of an earlier operating pressurized water reactor (PWR) plant cycle length analysis to account for differences between the design of IRIS and this operating PWR, are presented. The proposed methodology to resolve these maintenance-related barriers by the design process is described. The results of applying the methodology to two potential IRIS cycle length barriers, relief valve testing and emergency heat removal system testing, are presented

Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

Energy Technology Data Exchange (ETDEWEB)

Turinsky, Paul J., E-mail: turinsky@ncsu.edu [North Carolina State University, PO Box 7926, Raleigh, NC 27695-7926 (United States); Kothe, Douglas B., E-mail: kothe@ornl.gov [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6164 (United States)

2016-05-15

The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics “core simulator” based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL

Recent advances in severe accident technology - direct containment heating in advanced light water reactors

International Nuclear Information System (INIS)

Fontana, M.H.

1993-01-01

The issues affecting high-pressure melt ejection (HPME) and the consequential containment pressurization from direct containment heating (DCH), as they affect advanced light water reactors (ALWRs), specifically advanced pressurized water reactors (APWRs), were reviewed by the U.S. Department of Energy Advanced Reactor Severe Accident Program (ARSAP). Recommendations from ARSAP regarding the design of APWRs to minimize DCH are embodied within the Electric Power Research Institute ALWR Utility Requirements Document, which specifies (a) a large, strong containment; (b) an in-containment refueling water storage tank; (c) a reactor cavity configuration that minimizes energy transport to the containment atmosphere; and (d) a reactor coolant system depressurization system. Experimental and analytical efforts, which have focused on current-generation plants, and analyses for APWRs were reviewed. Although DCH is a subject of continuous research and considerable uncertainties remain, it is the judgment of the ARSAP that reactors complying with the recommended design requirements would have a low probability of early containment failure due to HPME and DCH

Potential of light water reactors for future nuclear power plants

International Nuclear Information System (INIS)

Gueldner, R.

2003-01-01

Energy consumption worldwide is going to increase further in the next few decades. Reliable supplies of electricity can be achieved only by centralized power plant structures. In this scenario, nuclear power plants are going to play a leading role as reliable and competitive plants, also under deregulated market conditions. Today, light water reactors have achieved a leading position, both technically and economically, contributing 85% to worldwide electricity generation in nuclear plants. They will continue to be a proven technology in power generation. In many countries, activities therefore are concentrated on extending the service life of plants beyond a period of forty years. New nuclear generating capacities are expected to be created and added from the end of this decade onward. Most of this capacity will be in light water reactors. The concepts of third-generation reactors will meet all economic and technical safety requirements of the 21st century and will offer considerable potential for further development. Probably some thirty years from now, fourth-generation nuclear power plants will be ready for commercial application. These plants will penetrate especially new sectors of the energy markets. Public acceptance of new nuclear power plants is not a matter of reactor lines, provided that safety requirements are met. The important issue is the management of radioactive waste. The construction of new nuclear power plants in Western Europe and North America mainly hinges on the ability to explain to the public that there is a need for new plants and that nuclear power is fundamental to assuring sustainable development. (orig.)

Nuclear powerplant standardization: light water reactors. Volume 2. Appendixes

International Nuclear Information System (INIS)

1981-06-01

This volume contains working papers written for OTA to assist in preparation of the report, NUCLEAR POWERPLANT STANDARDIZATION: LIGHT WATER REACTORS. Included in the appendixes are the following: the current state of standardization, an application of the principles of the Naval Reactors Program to commercial reactors; the NRC and standardization, impacts of nuclear powerplant standardization on public health and safety, descriptions of current control room designs and Duke Power's letter, Admiral Rickover's testimony, a history of standardization in the NRC, and details on the impact of standardization on public health and safety

International academic program in technologies of light-water nuclear reactors. Phases of development and implementation

International Nuclear Information System (INIS)

Geraskin, N I; Glebov, V B

2017-01-01

The results of implementation of European educational projects CORONA and CORONA II dedicated to preserving and further developing nuclear knowledge and competencies in the area of technologies of light-water nuclear reactors are analyzed. Present article addresses issues of design and implementation of the program for specialized training in the branch of technologies of light-water nuclear reactors. The systematic approach has been used to construct the program for students of nuclear specialties, which corresponding to IAEA standards and commonly accepted nuclear principles recognized in the European Union. Possibilities of further development of the international cooperation between countries and educational institutions are analyzed. Special attention is paid to e-learning/distance training, nuclear knowledge preservation and interaction with European Nuclear Education Network. (paper)

Status of the advanced boiling water reactor and simplified boiling water reactor

International Nuclear Information System (INIS)

Smith, P.F.

1992-01-01

This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

The industry/EPRI advanced light water reactor program

International Nuclear Information System (INIS)

Stahlkopf, K.E.; Noble, D.M.; Sugnet, W.R.; Bilan, W.J.

1986-01-01

For the United States nuclear power industry to remain viable, it must be prepared to meet the expected need for new generating capacity in the late 1990s with an improved reactor system. The best hope of meeting this requirement is with evolutionary changes in current LWR systems through system simplification and reevaluation of safety and operational design margins. The grid characteristics and the difficulty in raising capital for large projects indicate that smaller light water reactors (400 to 600 MWe) may play an important role the next generation

IAEA'S study on advanced applications of water cooled nuclear power plants

International Nuclear Information System (INIS)

Cleveland, J.; McDonald, A.; Rao, A.; )

2008-01-01

About one-fifth of the world's energy consumption is used for electricity generation, with nuclear power contributing approximately 15.2% of this electricity. However; most of the world's energy consumption is for heat and transportation. Nuclear energy has considerable potential to penetrate these energy sectors now served by fossil fuels that are characterized by price volatility and finite supply. Advanced applications of nuclear energy include seawater desalination, district heating, and heat for industrial processes. Nuclear energy also has potential to provide a near-term, greenhouse gas free, source of energy for transportation. These applications rely on a source of heat and electricity. Nuclear energy from water-cooled reactors, of course, is not unique in this sense. Indeed, higher temperature heat can be produced by burning natural gas and coal, or through the use of other nuclear technologies such as gas-cooled or liquid-metal-cooled reactors. Water-cooled reactors, however; are being deployed today while other reactor types have had considerably less operational and regulatory experience and will take still some time to be widely accepted in the market. Both seawater desalination and district heating with nuclear energy are well proven, and new seawater desalination projects using water-cooled reactors will soon be commissioned. Provision of process heat with nuclear energy can result in less dependence on fossil fuels and contribute to reductions of greenhouse gases. Importantly, because nuclear power produces base-load electricity at stable and predictable prices, it provides a greenhouse gas free source of electricity for transportation systems (trains and subways), and for electric and plug-in hybrid vehicles, and in the longer term nuclear energy could produce hydrogen for fuel cell vehicles, as well as for other components of a hydrogen economy. These advanced applications can play an important role in enhancing public acceptance of nuclear

Advanced light water reactor program at ABB-Combustion Engineering Nuclear Power

International Nuclear Information System (INIS)

Cahn, H.

1990-01-01

To meet the needs of Electric Utilities ordering nuclear power plants in the 1990s, ABB-Combustion Engineering is developing two designs which will meet EPRI consensus requirements and new licensing issues. The System 80 Plus design is an evolutionary pressurized water reactor plant modelled after the successful System 80 design in operation in Palo Verde and under construction in Korea. System Plus is currently under review by the US Nuclear Regulatory Commission with final design approval expected in 1991 and design certification in 1992. The Safe Integral Reactor (SIR) plant is a smaller facility with passive safety features and modular construction intended for design certification in the late 1990s. (author)

Consideration of important technical issues for advanced light water reactors

International Nuclear Information System (INIS)

Thadani, A.C.; Perch, R.L.

1993-01-01

Early in the design and review process of the Advanced Light Water Reactors (ALWR), the US Nuclear Regulatory Commission (NRC) in recognition of the importance of defense-in-depth focused its attention on lessons learned from the operating experience, research and other studies as well as addressing the challenges from severe accidents. The Commission issued the Policy Statement on Safety Goals for the Operations of Nuclear Power Plants on August 4, 1986. This policy statement focused on the risks to the public from nuclear power plant operations with the objective of establishing goals that broadly define an acceptable level of radiological risk that might be imposed on the public as a result of nuclear power plant operation. The Commission recognizes the importance of mitigating the consequences of a core-melt accident and continues to emphasize features such as containment and siting in less populated areas as integral parts of the defense-in-depth concept associated with its accident prevention and mitigation philosophy. In its Severe Accident Policy statement, the Commission expressed its expectation that vendors engage in designing new standard plants should address severe accidents during the design stage to take full advantage of insights gained by providing design features to further reduce the likelihood of severe accidents from occurring and, in the unlikely occurrence of a severe accident, mitigating their consequences. Incorporating insights and design features during the design phase can be cost effective when compared to modifications to existing plants. The staff has used this guidance to apply defense-in-depth philosophy in focusing attention on severe accident considerations. This paper discusses some of the key prevention and mitigation issues the NRC has focused its efforts, including emerging technologies being applied to new reactor designs

Development of the fuel-cycle costs in nuclear power stations with light-water reactors

International Nuclear Information System (INIS)

Brosch, R.; Moraw, G.; Musil, G.; Schneeberger, M.

1976-01-01

The authors investigate the fuel-cycle costs in nuclear power stations with light-water reactors in the Federal Republic of Germany in the years 1966 to 1976. They determine the effect of the price development for the individual components of the nuclear fuel cycle on the fuel-cycle costs averaged over the whole power station life. Here account is taken also of inflation rates and the change in the DM/US $ parity. In addition they give the percentage apportionment of the fuel-cycle costs. The authors show that real fuel-cycle costs for nuclear power stations with light-water reactors in the Federal Republic of Germany have risen by 11% between 1966 and 1976. This contradicts the often repeated reproach that fuel costs in nuclear power stations are rising very steeply and are no longer competitive. (orig.) [de

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.

Fuel cycle options for light water reactors and heavy water reactors. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1999-11-01

In the second half of the 20th century nuclear power has evolved from the research and development environment to an industry that supplies 16% of the world's electricity. By the end of 1997, over 8500 reactor-years of operating experience had been accumulated. Global environmental change, and the continuing increase in global energy supply required to provide increasing populations with an improving standard of living, make the contribution from nuclear energy even more important for the next century. For nuclear power to achieve its full potential and make its needed contribution, it must be safe, economical, reliable and sustainable. All of these factors can be enhanced by judicious choice and development of advanced fuel cycle options. The Technical Committee Meeting (TCM) on Fuel Cycle Options for Light Water Reactors and Heavy Water Reactors was hosted by Atomic Energy of Canada Limited (AECL) on behalf of the Canadian Government and was jointly conducted within the frame of activities of the IAEA International Working Group on Advanced Technologies for Light Water Reactors (IWG-LWR) and the IAEA International Working Group on Advanced Technologies for Heavy Water Reactors (IWG-HWR). The TCM provided the opportunity to have in-depth discussions on important technical topics which were highlighted in the International Symposium on Nuclear Fuel Cycle and Reactor Strategies: Adjusting to New Realities, held in Vienna, 3-6 June 1997. The main results and conclusions of the TCM were presented as input for discussion at the first meeting of the IAEA newly formed International Working Group on Fuel Cycle Options

Advanced nuclear reactor safety issues and research needs

International Nuclear Information System (INIS)

2002-01-01

On 18-20 February 2002, the OECD Nuclear Energy Agency (NEA) organised, with the co-sponsorship of the International Atomic Energy Agency (IAEA) and in collaboration with the European Commission (EC), a Workshop on Advanced Nuclear Reactor Safety Issues and Research Needs. Currently, advanced nuclear reactor projects range from the development of evolutionary and advanced light water reactor (LWR) designs to initial work to develop even further advanced designs which go beyond LWR technology (e.g. high-temperature gas-cooled reactors and liquid metal-cooled reactors). These advanced designs include a greater use of advanced technology and safety features than those employed in currently operating plants or approved designs. The objectives of the workshop were to: - facilitate early identification and resolution of safety issues by developing a consensus among participating countries on the identification of safety issues, the scope of research needed to address these issues and a potential approach to their resolution; - promote the preservation of knowledge and expertise on advanced reactor technology; - provide input to the Generation IV International Forum Technology Road-map. In addition, the workshop tried to link advancement of knowledge and understanding of advanced designs to the regulatory process, with emphasis on building public confidence. It also helped to document current views on advanced reactor safety and technology, thereby contributing to preserving knowledge and expertise before it is lost. (author)

PARs for combustible gas control in advanced light water reactors

International Nuclear Information System (INIS)

Hosler, J.; Sliter, G.

1997-01-01

This paper discusses the progress being made in the United States to introduce passive autocatalytic recombiner (PAR) technology as a cost-effective alternative to electric recombiners for controlling combustible gas produced in postulated accidents in both future Advanced Light Water Reactors (ALWRs) and certain U. S. operating nuclear plants. PARs catalytically recombine hydrogen and oxygen, gradually producing heat and water vapor. They have no moving parts and are self-starting and self-feeding, even under relatively cold and wet containment conditions. Buoyancy of the hot gases they create sets up natural convective flow that promotes mixing of combustible gases in a containment. In a non-inerted ALWR containment, two approaches each employing a combination of PARs and igniters are being considered to control hydrogen in design basis and severe accidents. In pre-inerted ALWRs, PARs alone control radiolytic oxygen produced in either accident type. The paper also discusses regulatory feedback regarding these combustible gas control approaches and describes a test program being conducted by the Electric Power Research Institute (EPRI) and Electricite de France (EdF) to supplement the existing PAR test database with performance data under conditions of interest to U.S. plants. Preliminary findings from the EPRI/EdF PAR model test program are included. Successful completion of this test program and confirmatory tests being sponsored by the U. S. NRC are expected to pave the way for use of PARs in ALWRs and operating plants. (author)

Development of Advanced High Uranium Density Fuels for Light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Blanchard, James [Univ. of Wisconsin, Madison, WI (United States); Butt, Darryl [Boise State Univ., ID (United States); Meyer, Mitchell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

2016-02-15

This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

Energy Technology Data Exchange (ETDEWEB)

Kristine Barrett; Shannon Bragg-Sitton

2012-09-01

The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

The United States Advanced Light Water reactor (USALWR) development program

International Nuclear Information System (INIS)

Stahlkopf, K.E.; Noble, D.M.; Devine, Jr.J.C.; Sugnet, W.R.

1987-01-01

For the United States Nuclear power industry to remain viable, it must be prepared to meet the expected need for a new generation capacity in the late 90s with an improved reactor system. The best hope of meeting this requirement is with revolutionary changes to current LWR systems through simplification and re-evaluation of safety and operational design margins. In addition, the grid characteristics and the difficulty in raising capital for large projects indicate the smaller light water reactors (600 MWe) may play an important role in the next generation. A cooperative and coordinated program between EPRI, U.S. DOE, the major architect engineers, nuclear steam supply vendors, and the NRC in the U.S. has been undertaken with four major goals in mind. (author)

The United States advanced light water reactor (USALWR) development program

International Nuclear Information System (INIS)

Stahlkopf, K.E.; Noble, D.M.; Devine, J.C. Jr.; Sugnet, W.R.

1987-01-01

For the United States Nuclear Power industry to remain viable, it must be prepared to meet the expected need for a new generation capacity in the late 90s with an improved reactor system. The best hope of meeting this requirement is with revolutionary changes to current LWR systems through simplification and re-evaluation of safety and operational design margins. In addition, the grid characteristics and the difficulty in raising capital for large projects indicate the smaller light water reactors (600 MWe) may play an important role in the next generation. A cooperative and coordinated program between EPRI, U.S. DOE, the major architect engineers, nuclear steam supply vendors, and the NRC in the U.S. has been undertaken with four major goals in mind

On the path to ordering standardized advanced light water reactors

International Nuclear Information System (INIS)

Sliter, G.E.

1997-01-01

The international Advanced Light Water Reactor (ALWR) program is specifying, designing, and certifying the next generation of nuclear power plants. Begun in the mid-1980's, the program is on track to permit ordering and construction of families of standardized plants at the start of the twenty-first century. ALWRs will be constructed only if they are economically competitive with alternative forms of electricity generation and are recognized as acceptable and favorable by the public, prospective owners, and investors. This paper first gives an overview of the major building blocks ensuring safe, reliable, and economic designs and the status of those designs. Next it lays out the path the industry has charted toward adopting the ALWR option and indicates the status of three key steps -- design certification, utility requirements, and first-of-a-kind engineering. Lastly, the paper focuses on one of the most important building blocks for ensuring economic viability -- life-cycle standardization. Among the topics are the definition and scope of standardization; its advantages and disadvantages; design team standardization plans that describe the desired or optimum degree of standardization and the processes used to achieve it; and the need for an agreement among all plant owners and operators for implementing and sustaining standardization in families of ALWRs. 10 refs., 5 figs

Single-earthquake design for piping systems in advanced light water reactors

International Nuclear Information System (INIS)

Terao, D.

1993-01-01

Appendix A to Part 100 of Title 10 of the Code of Federal Regulations (10 CFR Part 100) requires, in part, that all structures, systems, and components of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public shall be designed to remain functional and within applicable stress and deformation limits when subject to an operating basis earthquake (OBE). The US Nuclear Regulatory Commission (NRC) is proposing changes to Appendix A to Part 100 to redefine the OBE at a level such that its purpose can be satisfied without the need to perform explicit response analyses. Consequently, only the safe-shutdown earthquake (SSE) would be required for the seismic design of safety-related structures, systems and components. The purpose of this paper is to discuss the proposed changes to existing seismic design criteria that the NRC staff has found acceptable for implementing the proposed rule change in the design of safety-related piping systems in the advanced light water reactor (ALWR) lead plant. These criteria apply only to the ALWR lead plant design and are not intended to replace the seismic design criteria approved by the Commission in the licensing bases of currently operating facilities. Although the guidelines described herein have been proposed for use as a pilot program for implementing the proposed rule change specifically for the ALWR lead plant, the NRC staff expects that these guidelines will also be applied to other ALWRs

Advanced LWR Nuclear Fuel Cladding Development

International Nuclear Information System (INIS)

Bragg-Sitton, S.; Griffith, G.

2012-01-01

The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R and D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an 'accident tolerant' fuel system that would offer improved coping time under accident scenarios. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. (author)

NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

International Nuclear Information System (INIS)

1994-08-01

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Open-Source Integrated Design-Analysis Environment For Nuclear Energy Advanced Modeling & Simulation Final Scientific/Technical Report

Energy Technology Data Exchange (ETDEWEB)

O' Leary, Patrick [Kitware, Inc., Clifton Park, NY (United States)

2017-01-30

The framework created through the Open-Source Integrated Design-Analysis Environment (IDAE) for Nuclear Energy Advanced Modeling & Simulation grant has simplify and democratize advanced modeling and simulation in the nuclear energy industry that works on a range of nuclear engineering applications. It leverages millions of investment dollars from the Department of Energy's Office of Nuclear Energy for modeling and simulation of light water reactors and the Office of Nuclear Energy's research and development. The IDEA framework enhanced Kitware’s Computational Model Builder (CMB) while leveraging existing open-source toolkits and creating a graphical end-to-end umbrella guiding end-users and developers through the nuclear energy advanced modeling and simulation lifecycle. In addition, the work deliver strategic advancements in meshing and visualization for ensembles.

Advanced Light Water Reactor Plants System 80+trademark Design Certification Program

International Nuclear Information System (INIS)

1993-01-01

The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+trademark during the US government's 1993 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW t (1350 MWe) Pressurized Water Reactor (PWR). The design consists of an essentially complete plant. It is based on evolutionary improvements to the Standardized System 80 nuclear steam supply system in operation at Palo Verde Units 1, 2, and 3, and the Duke Power Company P-81 balance-of-plant (BOP) that was designed and partially constructed at the Cherokee plant site. The System 80/P-81 original design has been substantially enhanced to increase conformance with the EPRI ALWR Utility Requirements Document (URD). Some design enhancements incorporated in the System 80+ design are included in the four units currently under construction in the Republic of Korea. These units form the basis of the Korean standardization program. The full System 80+ standard design has been offered to the Republic of China, in response to their recent bid specification. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was submitted to the NRC and a Draft Safety Evaluation Report was issued by the NRC in October 1992. CESSAR-DC contains the technical basis for compliance with the EPRI URD for simplified emergency planning. The Nuclear Steam Supply System (NSSS) is the standard ABB-Combustion Engineering two-loop arrangement with two steam generators, two hot legs and four cold legs each with a reactor coolant pump. The System 80+ standard plant includes a sperical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual containment

Advanced Light Water Reactor Plants System 80+trademark Design Certification Program

International Nuclear Information System (INIS)

Davis, G.A.

1992-01-01

Since 1985, ABB Combustion Engineering Nuclear Power (CENP) and Duke Engineering ampersand Services, Inc. (DE ampersand S) have been developing the next generation of pressurized water reactor (PWR) plant for worldwide deployment. The goal is to make available a pre-licensed, standardized plant design that can satisfy the need for a reliable and economic supply of electricity for residential, commercial and industrial use. To ensure that such a design is available when needed, it must be based on proven technology and established licensing criteria. These requirements dictate development of nuclear technology that is advanced, yet evolutionary in nature. This has been achieved with the System 80+ Standard Plant Design

Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

Science.gov (United States)

Turinsky, Paul J.; Kothe, Douglas B.

2016-05-01

The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics ;core simulator; based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL's evolving M

Cost-estimate guidelines for advanced nuclear power technologies

International Nuclear Information System (INIS)

Delene, J.G.; Hudson, C.R.

1993-01-01

Various advanced power plant concepts are currently under development. These include several advanced light water reactors as well as the modular high-temperature gas-cooled reactor and the advanced liquid-metal reactor. One measure-of the attractiveness of a new concept is cost. Invariably, the cost of a new type of power plant will be compared with other alternative forms of electric generation. In order to make reasonable comparative assessments of competing technologies, consistent ground rules and assumptions must be applied when developing cost estimates. This paper describes the cost-estimate guidelines developed by Oak Ridge National Laboratory for the U.S. Department of Energy (DOE) to be used in developing cost estimates for the advanced nuclear reactors and how these guidelines relate to the DOE cost verification process

Materials technologies of light water reactors

International Nuclear Information System (INIS)

Begley, R.

1984-01-01

Satisfactory materials performance is a key element in achieving reliable operation of light water reactors. Outstanding performance under rigorous operational conditions has been exhibited by pressure boundary components, core internals, fuel cladding, and other critical components of these systems. Corrosion and stress corrosion phenomena have, however, had an impact on plant availability, most notably relating to pipe cracking in BWR systems and steam generator corrosion in PWR systems. These experiences have stimulated extensive development activities by the nuclear industry in improved NDE techniques, investigation of corrosion phenomena, as well as improved materials and repair processes. This paper reviews key materials performance aspects of light water reactors with particular emphasis on the progress which has been made in modeling of corrosion phenomena, control of the plant operating environment, advanced material development, and application of sophisticated repair procedures. Implementation of this technology provides the basis for improved plant availability

NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER provides a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Neutron fluence determination for light water reactor pressure vessels

International Nuclear Information System (INIS)

Gold, R.

1994-01-01

A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

International Nuclear Information System (INIS)

1994-08-01

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Materials for advanced water cooled reactors

International Nuclear Information System (INIS)

1992-09-01

The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

Prospects for and problems of using light-water supercritical-pressure coolant in nuclear reactors in order to increase the efficiency of the nuclear fuel cycle

International Nuclear Information System (INIS)

Alekseev, P. N.; Semchenkov, Yu. M.; Sedov, A. A.; Subbotin, S. A.; Chibinyaev, A. V.

2011-01-01

Trends in the development of the power sector of the Russian and world power industries both at present time and in the near future are analyzed. Trends in the rise of prices for reserves of fossil and nuclear fuels used for electricity production are compared. An analysis of the competitiveness of electricity production at nuclear power plants as compared to the competitiveness of electricity produced at coal-fired and natural-gas-fired thermal power plants is performed. The efficiency of the open nuclear fuel cycle and various versions of the closed nuclear fuel cycle is discussed. The requirements on light-water reactors under the scenario of dynamic development of the nuclear power industry in Russia are determined. Results of analyzing the efficiency of fuel utilization for various versions of vessel-type light-water reactors with supercritical coolant are given. Advantages and problems of reactors with supercritical-pressure water are listed.

Report from the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies

International Nuclear Information System (INIS)

Hallbert, Bruce P.; Persensky, J.J.; Smidts, Carol; Aldemir, Tunc; Naser, Joseph

2009-01-01

The Light Water Reactor Sustainability (LWRS) Program is a research and development (R and D) program sponsored by the U.S. Department of Energy (DOE). The program is operated in close collaboration with industry R and D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of Nuclear Power Plants that are currently in operation. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. Advanced instruments and control (I and C) technologies are needed to support the safe and reliable production of power from nuclear energy systems during sustained periods of operation up to and beyond their expected licensed lifetime. This requires that new capabilities to achieve process control be developed and eventually implemented in existing nuclear assets. It also requires that approaches be developed and proven to achieve sustainability of I and C systems throughout the period of extended operation. The strategic objective of the LWRS Program Advanced Instrumentation, Information, and Control Systems Technology R and D pathway is to establish a technical basis for new technologies needed to achieve safety and reliability of operating nuclear assets and to implement new technologies in nuclear energy systems. This will be achieved by carrying out a program of R and D to develop scientific knowledge in the areas of: (1) Sensors, diagnostics, and prognostics to support characterization and prediction of the effects of aging and degradation phenomena effects on critical systems, structures, and components (SSCs); (2) Online monitoring of SSCs and active components, generation of information, and methods to analyze and employ online monitoring information; (3) New methods for visualization, integration, and information use to enhance state awareness and leverage expertise to achieve safer, more readily available

Report from the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies

Energy Technology Data Exchange (ETDEWEB)

Bruce P. Hallbert; J. J. Persensky; Carol Smidts; Tunc Aldemir; Joseph Naser

2009-08-01

The Light Water Reactor Sustainability (LWRS) Program is a research and development (R&D) program sponsored by the U.S. Department of Energy (DOE). The program is operated in close collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of Nuclear Power Plants that are currently in operation. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. Advanced instruments and control (I&C) technologies are needed to support the safe and reliable production of power from nuclear energy systems during sustained periods of operation up to and beyond their expected licensed lifetime. This requires that new capabilities to achieve process control be developed and eventually implemented in existing nuclear assets. It also requires that approaches be developed and proven to achieve sustainability of I&C systems throughout the period of extended operation. The strategic objective of the LWRS Program Advanced Instrumentation, Information, and Control Systems Technology R&D pathway is to establish a technical basis for new technologies needed to achieve safety and reliability of operating nuclear assets and to implement new technologies in nuclear energy systems. This will be achieved by carrying out a program of R&D to develop scientific knowledge in the areas of: • Sensors, diagnostics, and prognostics to support characterization and prediction of the effects of aging and degradation phenomena effects on critical systems, structures, and components (SSCs) • Online monitoring of SSCs and active components, generation of information, and methods to analyze and employ online monitoring information • New methods for visualization, integration, and information use to enhance state awareness and leverage expertise to achieve safer, more readily available electricity generation

International light water nuclear fuel fabrication supply. Are fabrication services assured?

International Nuclear Information System (INIS)

Rothwell, Geoffrey

2010-01-01

This paper examines the cost structure of fabricating light water reactor (LWR) fuel with low-enriched uranium (LEU, with less than 5% enrichment). The LWR-LEU fuel industry is decades old, and (except for the high entry cost of designing and licensing a fuel fabrication facility and its fuel), labor and additional fabrication lines can be added at Nth-of-a-Kind cost to the maximum capacity allowed by a site license. The industry appears to be competitive: nuclear fuel fabrication capacity is assured with many competitors and reasonable prices. However, nuclear fuel assurance has become an important issue for nations now to considering new nuclear power plants. To provide this assurance many proposals equate 'nuclear fuel banks' (which would require fuel for specific reactors) with 'LEU banks' (where LEU could be blended into nuclear fuel with the proper enrichment) with local fuel fabrication. The policy issues (which are presented, but not answered in this paper) become (1) whether the construction of new nuclear fuel fabrication facilities in new nuclear power nations could lead to the proliferation of nuclear weapons, and (2) whether nuclear fuel quality can be guaranteed under current industry arrangements, given that fuel failure at one reactor can lead to forced shutdowns at many others. (author)

Energy-analysis of the total nuclear energy cycle based on light water reactors

International Nuclear Information System (INIS)

Kistemaker, J.

1975-01-01

The energy economy of the total nuclear energy cycle is investigated. Attention is paid to the importance of fossil fuel saving by using nuclear energy. The energy analysis is based on the construction and operation of power plants with an electric output of 1000MWe. Light water moderated reactors with a 2.7 - 3.2% enriched uranium core are considered. Additionally, the whole fuel cycle including ore winning and refining, enrichment and fuel element manufacturing and reprocessing has been taken into account. Neither radioactive waste storage problems nor safety problems related to the nuclear energy cycle and safeguarding have been dealt with, as exhaustive treatments can be found elswhere

Westinghouse AP600 advanced nuclear plant design

International Nuclear Information System (INIS)

Gangloff, W.

1999-01-01

As part of the cooperative US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) Program and the Electric Power Research Institute (EPRI), the Westinghouse AP600 team has developed a simplified, safe, and economic 600-megawatt plant to enter into a new era of nuclear power generation. Designed to satisfy the standards set by DOE and defined in the ALWR Utility Requirements Document (URD), the Westinghouse AP600 is an elegant combination of innovative safety systems that rely on dependable natural forces and proven technologies. The Westinghouse AP600 design simplifies plant systems and significant operation, inspections, maintenance, and quality assurance requirements by greatly reducing the amount of valves, pumps, piping, HVAC ducting, and other complex components. The AP600 safety systems are predominantly passive, depending on the reliable natural forces of gravity, circulation, convection, evaporation, and condensation, instead of AC power supplies and motor-driven components. The AP600 provides a high degree of public safety and licensing certainty. It draws upon 40 years of experience in light water reactor components and technology, so no demonstration plant is required. During the AP600 design program, a comprehensive test program was carried out to verify plant components, passive safety systems components, and containment behavior. When the test program was completed at the end of 1994, the AP600 became the most thoroughly tested advanced reactor design ever reviewed by the US Nuclear Regulatory Commission (NRC). The test results confirmed the exceptional behavior of the passive systems and have been instrumental in facilitating code validations. Westinghouse received Final Design Approval from the NRC in September 1998. (author)

Passive and inherent safety technologies for light-water nuclear reactors

International Nuclear Information System (INIS)

Forsberg, C.W.

1990-07-01

Passive/inherent safety implies a technical revolution in our approach to nuclear power safety. This direction is discussed herein for light-water reactors (LWRs) -- the predominant type of power reactor used in the world today. At Oak Ridge National Laboratory (ORNL) the approach to the development of passive/inherent safety for LWRs consists of four steps: identify and quantify safety requirements and goals; identify and quantify the technical functional requirements needed for safety; identify, invent, develop, and quantify technical options that meet both of the above requirements; and integrate safety systems into designs of economic and reliable nuclear power plants. Significant progress has been achieved in the first three steps of this program. The last step involves primarily the reactor vendors. These activities, as well as related activities worldwide, are described here. 27 refs., 7 tabs

Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

D. E. Shropshire

2009-01-01

The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

International Nuclear Information System (INIS)

Shropshire, D.E.

2009-01-01

The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program's understanding of the cost drivers that will determine nuclear power's cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-irradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

Comparison between Nuclear Data Libraries of Different Density of Data for H in Light Water

International Nuclear Information System (INIS)

Torres, Lourdes; Gillette, Victor

2003-01-01

We introduce the results of comparison between nuclear data libraries at different density of data.Nuclear data libraries were produced for hydrogen (H) in light water at different density of data.These libraries were produced using the NJOY nuclear data processing system.With this code we produce pointwise cross sections and related quantities, in the ENDF format, and in the ACE format for MCNP.Experimental neutron spectrum was compared with MCNP4C simulations, based on the produced libraries and calculation time

Two-phase coolant pump model of pressurized light water nuclear reactors

International Nuclear Information System (INIS)

Santos, G.A. dos; Freitas, R.L.

1990-01-01

The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The homologous curves set up the complete performance of the pump and are input for accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

Process for superheating the steam generated by a light water nuclear reactor

International Nuclear Information System (INIS)

Vakil, H.B.; Brown, D.H.

1976-01-01

A process is submitted for superheating the pressurised steam generated in a light water nuclear reactor in which the steam is brought to 340 0 C at least. This superheated steam is used to operate a turbo-generator unit. The characteristic of the process is that an exothermal chemical reaction is used to generate the heat utilised during the superheating stage. The chemical reaction is a mechanisation, oxidation-reduction or hydrogenation reaction [fr

Commercializing the next generation: the AP600 advanced simplified nuclear power plant

International Nuclear Information System (INIS)

Bruschi, H.J.

1994-01-01

Today, government and industry are working together on advanced nuclear power plant designs that take advantage of valuable lessons learned from the experience to date and promise to reconcile the demands of economic expansion with the laws of environmental protection. In the U.S., the Department of Energy (DOE) and the Electric Power Research Institute (EPRI) initiated a design certification program in 1989 to develop and commercialize advanced light water reactors (ALWRs) for the next round of power plant construction. Advanced, simplified technology is one approach under development to end the industry's search for a simpler, more forgiving, and less costly reactor. As part of this program, Westinghouse is developing the AP600, a new standard 600 MWe advanced, simplified plant. The design strikes a balance between the use of proven technology and new approaches. The result is a greatly streamlined plant that can meet safety regulations and reliability requirements, be economically competitive, and promote broader public confidence in nuclear energy. 1 fig

Water cooled metal optics for the Advanced Light Source

International Nuclear Information System (INIS)

McKinney, W.R.; Irick, S.C.; Lunt, D.L.J.

1991-01-01

The program for providing water cooled metal optics for the Advanced Light Source at Berkeley is reviewed with respect to fabrication and metrology of the surfaces. Materials choices, surface figure and smoothness specifications, and metrology systems for measuring the plated metal surfaces are discussed. Results from prototype mirrors and grating blanks will be presented, which show exceptionally low microroughness and mid-period error. We will briefly describe out improved version of the Long Trace Profiler, and its importance to out metrology program. We have completely redesigned the mechanical, optical and computational parts of the profiler system with the cooperation of Peter Takacs of Brookhaven, Continental Optical, and Baker Manufacturing. Most important is that one of our profilers is in use at the vendor to allow testing during fabrication. Metrology from the first water cooled mirror for an ALS beamline is presented as an example. The preplating processing and grinding and polishing were done by Tucson Optical. We will show significantly better surface microroughness on electroless nickel, over large areas, than has been reported previously

Overview of Light Hydrogen-Based Low Energy Nuclear Reactions

Science.gov (United States)

Miley, George H.; Shrestha, Prajakti J.

This paper reviews light water and hydrogen-based low-energy nuclear reactions (LENRs) including the different methodologies used to study these reactions and the results obtained. Reports of excess heat production, transmutation reactions, and nuclear radiation emission are cited. An aim of this review is to present a summary of the present status of light water LENR research and provide some insight into where this research is heading.

Advanced fuel assemblies for economic and flexible operation of light water reactors

International Nuclear Information System (INIS)

Urban, P.; Bender, D.

2001-01-01

Increasing competition in the electricity market sets up a corresponding competition between the different electricity producing technologies. This makes further improvements in the economics of nuclear power generation a vital item for the future of nuclear energy. Though the costs for development, design and fabrication of fuel assemblies contribute only about 10% to the fuel cycle costs, the design and the performance of the fuel assemblies considerably influences total electricity generation cost. By the recent creation of Framatome ANP the nuclear activities of Framatome and Siemens were combined into one company. In the past, both had made considerable achievements in the development of fuel assemblies and related services supporting the goal of safe and economic electricity generation by light water reactors. The examples described in this paper cover former Siemens products and experience. In the future, our combined experience bases will be an ideal platform to offer further substantial improvements to our customers. (author)

Assessment of United States industry structural codes and standards for application to advanced nuclear power reactors: Appendices. Volume 2

International Nuclear Information System (INIS)

Adams, T.M.; Stevenson, J.D.

1995-10-01

Throughout its history, the USNRC has remained committed to the use of industry consensus standards for the design, construction, and licensing of commercial nuclear power facilities. The existing industry standards are based on the current class of light water reactors and as such may not adequately address design and construction features of the next generation of Advanced Light Water Reactors and other types of Advanced Reactors. As part of their on-going commitment to industry standards, the USNRC commissioned this study to evaluate US industry structural standards for application to Advanced Light Water Reactors and Advanced Reactors. The initial review effort included (1) the review and study of the relevant reactor design basis documentation for eight Advanced Light Water Reactors and Advanced Reactor Designs, (2) the review of the USNRCs design requirements for advanced reactors, (3) the review of the latest revisions of the relevant industry consensus structural standards, and (4) the identification of the need for changes to these standards. The results of these studies were used to develop recommended changes to industry consensus structural standards which will be used in the construction of Advanced Light Water Reactors and Advanced Reactors. Over seventy sets of proposed standard changes were recommended and the need for the development of four new structural standards was identified. In addition to the recommended standard changes, several other sets of information and data were extracted for use by USNRC in other on-going programs. This information included (1) detailed observations on the response of structures and distribution system supports to the recent Northridge, California (1994) and Kobe, Japan (1995) earthquakes, (2) comparison of versions of certain standards cited in the standard review plan to the most current versions, and (3) comparison of the seismic and wind design basis for all the subject reactor designs

Assuring PSA technical adequacy for new advanced light water reactor designs

International Nuclear Information System (INIS)

Lutz, R.J.; Detar, H.L.; Schneider, R.E.

2012-01-01

The Probabilistic Safety Assessment (PSA) for an Advanced Light Water Reactor (ALWR) must exhibit a high level of technical adequacy, or technical quality, in order to be used as a reliable tool for making risk informed decisions concerning design and eventual operation of the plant. During the design phase, decisions on some design features may use the PSA as an input. Also, the PSA may be used as input to other operational decisions during plant design and construction including the development of procedures, development of technical specification limiting conditions for operation and scheduling of preventive maintenance activities. For the existing fleet of light water reactors (LWRs), PSA technical adequacy can be judged from wide ranging acceptance criteria such as the PRA (Probabilistic Risk Assessment) Standard in the United States of America that was developed jointly by the American Society of Mechanical Engineers (ASME) and the American Nuclear Society (ANS). However, the requirements for PRA technical adequacy in this PRA Standard assumes that the plant is built and has operation experience. Some of the requirements cannot be met for ALWRs in the design or construction phase and with no operational history. Key elements of a high level of technical adequacy include procedures, operator interviews, plant walk-downs and equipment reliability histories. The ability to include these key elements into the ALWR PSA to improve technical adequacy will progress as the ALWR progresses from the design stage through the construction stage and finally to the fuel load / pre-operational stage. As the technical adequacy becomes more robust, more confidence can be placed on risk-informed decisions that are made with the PSA. To assist in using the PSA as input to design and operational decisions in the design and construction stages of an ALWR, an addition to the ASME/ANS PRA Standard is being developed. The intent of this addition to the Standard is to provide

Nondestructive examination (NDE) Reliability for Inservice Inspection of Light Water Reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Good, M.S.; Heasler, P.G.; Hockey, R.L.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

1992-07-01

The Evaluation and Improvement of NDE reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties

Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

Use of the modular modeling system in the design of the Penn State Advanced Light Water Reactor

International Nuclear Information System (INIS)

Smith, K.A.

1988-12-01

This study involves the design and subsequent transient analysis of the Penn State Advanced Light Water Reactor (PSU ALWR). The performance of the PSU ALWR is evaluated during small step changes in power and a turbine trip from full power without scram. The Modular Modeling System (MMS), developed by Babcock and Wilcox under a contract from the Electric Power Research Institute (EPRI), is a computer code designed for the simulation of nuclear and fossil power plants. MMS uses preprogrammed modules to represent specific power plant components such as pipes, pumps, steam generators, and a nuclear reactor. These components can then be connected in any manner the user desires providing certain simple interconnection rules are followed. In this study, MMS is used to develop computer models of both the PSU ALWR and a conventional PWR operating at the same power level. These models are then subjected to the transients mentioned above to evaluate the ability of the letdown-injection system to maintain primary system pressure. The transient response of the PSU ALWR and conventional PWR MMS models were compared to each other and whenever possible to actual plant transient data. 14 refs., 29 figs., 5 tabs

Determination of heavy water in heavy water - light water mixtures

International Nuclear Information System (INIS)

Sanhueza M, A.

1986-01-01

A description about experimental methodology to determine isotopic composition of heavy water - light water mixtures is presented. The employed methods are Nuclear Magnetic Resonance Spectroscopy, for measuring heavy water concentrations from 0 to 100% with intervals of 10% approx., and mass Spectrometry, for measuring heavy water concentrations from 0.1 to 1% with intervals of 0.15% approx., by means of an indirect method of Dilution. (Author)

Recent development of nuclear power in Japan and instrumentation and control system and control room equipment for advanced light water reactors

International Nuclear Information System (INIS)

Wakayama, N.

1992-01-01

This paper was provided for the 13th IAEA/IWG-NPPCI Meeting and aims to introduce an outline of recent development of nuclear power in Japan and some topics in the field of nuclear power plant control and instrumentation. Forty units of nuclear power plants are in operation in Japan and five units of BWRs and six PWRs are under construction. Construction of prototype FBR Monju have almost completed an construction of High-Temperature Engineering Test Reactor, HTTR, started in March 1991. In parallel of those, extensive effort has been carried out to develop the third generation LWRs which are called Advanced BWR (ABWR) and Advanced PWR (APWR). Two Advanced BWRs are under safety review for construction. Instrumentation and control system of these Advanced LWRs adopts integrated digital I and C system, optical multiplexing signal transmission, fault tolerant control systems and software logic for reactor protection and safety systems and enhances plant control performance and provides human-friendly operation and maintenance environments. Main control room of these Advanced LWRs, comprised with large display panels and advanced console, has special futures such as one-man sit-down operation, human friendly man-machine interface, high level automation in operation and maintenance. (author). 7 refs, 9 figs, 1 tab

Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

Energy Technology Data Exchange (ETDEWEB)

Unal, Cetin [Los Alamos National Laboratory; Pasamehmetoglu, Kemal [IDAHO NATIONAL LAB; Carmack, Jon [IDAHO NATIONAL LAB

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R & D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

International Nuclear Information System (INIS)

Unal, Cetin; Pasamehmetoglu, Kemal; Carmack, Jon

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R and D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

Qualification issues associated with the use of advanced instrumentation and control systems hardware in nuclear power plants

International Nuclear Information System (INIS)

Korsah, K.; Antonescu, C.

1993-01-01

The instrumentation and control (I ampersand C) systems in advanced reactors will make extensive use of digital controls, microprocessors, multiplexing, and Tiber-optic transmission. Elements of these advances in I ampersand C have been implemented on some current operating plants. However, the widespread use of the above technologies, as well as the use of artificial intelligence with minimum reliance on human operator control of reactors, highlights the need to develop standards for qualifying I ampersand C used in the next generation of nuclear power plants. As a first step in this direction, the protection system I ampersand C for present-day plants was compared to that proposed for advanced light water reactors (ALWRs). An evaluation template was developed by assembling a configuration of a safety channel instrument string for a generic ALWR, then comparing the impact of environmental stressors on that string to their effect on an equivalent instrument string from an existing light water reactor. The template was then used to address reliability issues for microprocessor-based protection systems. Standards (or lack thereof) for the qualification of microprocessor-based safety I ampersand C systems were also identified. This approach addresses in part issues raised in Nuclear Regulatory Commission policy document SECY-91-292. which recognizes that advanced I ampersand C systems for the nuclear industry are ''being developed without consensus standards, as the technology available for design is ahead of the technology that is well understood through experience and supported by application standards.''

Advanced modelling and numerical strategies in nuclear thermal-hydraulics

International Nuclear Information System (INIS)

Staedtke, H.

2001-01-01

The first part of the lecture gives a brief review of the current status of nuclear thermal hydraulics as it forms the basis of established system codes like TRAC, RELAP5, CATHARE or ATHLET. Specific emphasis is given to the capabilities and limitations of the underlying physical modelling and numerical solution strategies with regard to the description of complex transient two-phase flow and heat transfer conditions as expected to occur in PWR reactors during off-normal and accident conditions. The second part of the lecture focuses on new challenges and future needs in nuclear thermal-hydraulics which might arise with regard to re-licensing of old plants using bestestimate methodologies or the design and safety analysis of Advanced Light Water Reactors relying largely on passive safety systems. In order to meet these new requirements various advanced modelling and numerical techniques will be discussed including extended wellposed (hyperbolic) two-fluid models, explicit modelling of interfacial area transport or higher order numerical schemes allowing a high resolution of local multi-dimensional flow processes.(author)

Nuclear synergism of the light elements

International Nuclear Information System (INIS)

Harms, A.A.

1983-05-01

Some basic issues concerning accelerator initiated and fusion sustained nuclear energy systems are examined. For this purpose we identify selected nuclear fusion reactions characterized by a variable ion-to-neutron content and explore their intrinsic couplings and regenerative features. These are then related to particular systems concepts which emphasize fusion physics and accelerator technology. It is concluded that several light-element reaction systems possess appealing and interesting properties and can further be associated with selected advanced nuclear technologies. Their eventual implementation as nuclear energy systems requires further research in fusion physics, accelerator technology and mathematical physics. Because of the substantial potential benefits of such nuclear energy systems, it is concluded that research in this area should be pursued with much vigour. (orig.)

Advanced light water reactor plants System 80+trademark design certification program. Annual progress report, October 1, 1994 - September 30, 1995

International Nuclear Information System (INIS)

1998-01-01

The purpose of this report is to provide the status of the progress that was made towards Design Certification of System 80+trademark during the US government's 1995 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2, and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems

Advanced light water reactor plants System 80+trademark design certification program. Annual progress report, October 1, 1995 - September 30, 1996

International Nuclear Information System (INIS)

1996-01-01

The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+trademark during the US government's 1996 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2 and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems

Development of light water reactors and subjects for hereafter

International Nuclear Information System (INIS)

Murao, Yoshio

1995-01-01

As for light water reactors, the structure is relatively simple, and the power plants of large capacity can be realized easily, therefore, they have been used for long period as main nuclear reactors. During that period, the accumulation of experiences on the design, manufacture, operation, maintenance and regulation of light water has become enormous, and in Japan, the social base for maintaining and developing light water reactor technologies has been prepared sufficiently. If the nuclear power generation using seawater uranium is considered, the utilization of uranium for light water reactor technologies can become the method of producing the own energy for Japan. As the factors that threaten the social base of light water reactor technologies, there are a the lowering of the desire to promote light water reactors, the effect of secular deterioration, the price rise of uranium resources, the effect of plutonium accumulation, the effect of the circumstances in developing countries and the sure recruiting of engineers. The construction and the principle of working of light water reactors and the development of light water reactors hereafter, for example, the improvement on small scale and the addition of new technology resulting in cost reduction and the lowering of the quality requirement for engineers, the improvement of core design, the countermeasures by design to serious accidents and others are described. (K.I.)

Radiation and physical protection challenges at advanced nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Pickett, Susan E.

2008-01-01

Full text: The purpose of this study is to examine challenges and opportunities for radiation protection in advanced nuclear reactors and fuel facilities proposed under the Generation IV (GEN IV) initiative which is examining and pursuing the exploration and development of advanced nuclear science and technology; and the Global Nuclear Energy Partnership (GNEP), which seeks to develop worldwide consensus on enabling expanded use of economical, carbon-free nuclear energy to meet growing energy demand. The International Energy Agency projects nuclear power to increase at a rate of 1.3 to 1.5 percent a year over the next 20 years, depending on economic growth. Much of this growth will be in Asia, which, as a whole, currently has plans for 40 new nuclear power plants. Given this increase in demand for new nuclear power facilities, ranging from light water reactors to advanced fuel processing and fabrication facilities, it is necessary for radiation protection and physical protection technologies to keep pace to ensure both worker and public health. This paper is based on a review of current initiatives and the proposed reactors and facilities, primarily the nuclear fuel cycle facilities proposed under the GEN IV and GNEP initiatives. Drawing on the Technology Road map developed under GEN IV, this work examines the potential radiation detection and protection challenges and issues at advanced reactors, including thermal neutron spectrum systems, fast neutron spectrum systems and nuclear fuel recycle facilities. The thermal neutron systems look to improve the efficiency of production of hydrogen or electricity, while the fast neutron systems aim to enable more effective management of actinides through recycling of most components in the discharged fuel. While there are components of these advanced systems that can draw on the current and well-developed radiation protection practices, there will inevitably be opportunities to improve the overall quality of radiation

Regulation of nuclear power: the case of the light water reactor

International Nuclear Information System (INIS)

Rolph, E.

1977-06-01

This report is one of a series of documents that trace the history of the development and commercialization of the light water reactor in the expectation that a better appreciation of the development and commercialization process of this complex technology could be instructive in understanding the regulatory and economic obstacles currently slowing diffusion of that technology and the problems that may be encountered in similar large-scale, high-technology development projects. This regulatory history of the Atomic Energy Commission chronicles the significant events between 1954, when the AEC was given responsibility for regulating nuclear power plants, and 1974, when the AEC was absorbed by the Energy Research and Development Administration and the Nuclear Regulatory Commission. It identifies the origins of regulatory problems that have arisen during the period and describes how the AEC dealt with them. The history is based, for the most part, on primary source documents: hearings, news reports, and AEC documents

Qualification issues associated with the use of advanced instrumentation and control systems hardware in nuclear power plants

International Nuclear Information System (INIS)

Korsah, K.; Antonescu, C.

1993-01-01

The instrumentation and control (I ampersand C) systems in advanced reactors will make extensive use of digital controls, microprocessors, multiplexing, and fiber-optic transmission. Elements of these advances in I ampersand C have been implemented on some current operating plants. However, the widespread use of the above technologies, as well as the use of artificial intelligence with minimum reliance on human operator control of reactors, highlights the need to develop standards for qualifying I ampersand C used in the next generation of nuclear power plants. As a first step in this direction, the protection system I ampersand C for present-day plants was compared to that proposed for advanced light water reactors (ALWRs). An evaluation template was developed by assembling a configuration of a safety channel instrument string for a generic ALWR, then comparing the impact of environmental stressors on that string to their effect on an equivalent instrument string from an existing light water reactor. The template was then used to address reliability issues for microprocessor-based protection systems. Standards (or lack thereof) for the qualification of microprocessor-based safety I ampersand C systems were also identified. This approach addresses in part issues raised in Nuclear Regulatory Commission policy document SECY-91-292, which recognizes that advanced I ampersand C systems for the nuclear industry are open-quotes being developed without consensus standards, as the technology available for design is ahead of the technology that is well understood through experience and supported by application standards.close quotes

Advanced nuclear reactor and nuclear fusion power generation

International Nuclear Information System (INIS)

2000-04-01

This book comprised of two issues. The first one is a advanced nuclear reactor which describes nuclear fuel cycle and advanced nuclear reactor like liquid-metal reactor, advanced converter, HTR and extra advanced nuclear reactors. The second one is nuclear fusion for generation energy, which explains practical conditions for nuclear fusion, principle of multiple magnetic field, current situation of research on nuclear fusion, conception for nuclear fusion reactor and economics on nuclear fusion reactor.

LightSavers : accelerating advanced outdoor lighting

Energy Technology Data Exchange (ETDEWEB)

Purcell, B.; Pickering, M.

2010-01-15

This paper provided an update to the Toronto Atmospheric Fund (TAF) LightSavers program. The program was designed to accelerate market transformation for light emitting diode (LED) and advanced lighting management systems in outdoor lighting applications. It is expected that the program will result in significant electricity savings and emissions reductions within the Greater Toronto Area (GTA) and other Ontario municipalities. The first phase of the program established advanced outdoor lighting pilot programs in parking lots, garages, and pathway lighting applications that were guided by a common monitoring protocol to ensure useful and reliable assessment of the pilot programs. The TAF has since developed a strategy to strengthen public understanding and support for the use of advanced lighting, and continues to address policy issues that may impact the future of LED lighting programs. The TAF has also activated an electronic newsletter, delivered public workshops, and has been represented at several conferences. A working partnership has been established with Toronto Hydro Energy Services. Five pilot sites have been installed and have begun to provide monitoring data. Details of the pilot programs were provided. 16 figs.

Current status of materials development of nuclear fuel cladding tubes for light water reactors

International Nuclear Information System (INIS)

Duan, Zhengang; Yang, Huilong; Satoh, Yuhki; Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie; Abe, Hiroaki

2017-01-01

Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

Current status of materials development of nuclear fuel cladding tubes for light water reactors

Energy Technology Data Exchange (ETDEWEB)

Duan, Zhengang, E-mail: duan_zg@imr.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Yang, Huilong [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Satoh, Yuhki [Institute for Materials Research, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Abe, Hiroaki, E-mail: abe.hiroaki@n.t.u-tokyo.ac.jp [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan)

2017-05-15

Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

Nuclear technologies

International Nuclear Information System (INIS)

Toyama, Makoto; Hamasaki, Manabu; Kobayashi, Masahiko; Hoshide, Akihiko; Katayama, Kimio; Nozawa, H.; Karigome, Satoshi

2010-01-01

In recent days, energy security is becoming a major global concern and it has been recognized that a major reduction in greenhouse-gas emissions is required to combat climate change. Considerable expansion and new introduction of nuclear power generation are currently being planned and considered for the further in various parts of the world. Nuclear technologies of the latest 10 years in Japan were reviewed with their characteristics, advancement and future perspective. Steady efforts have been made to construct new nuclear power stations with computer-aided engineering system and modular and prefabricated structures, extend the interval of periodic inspections under the new inspection system that should improve both safety and reliability, implement advanced measures against aging and develop the next-generation light water reactors including a medium small reactor. Export of nuclear power plants has been promoted with international business alliance or cooperation. Activities to close nuclear fuel cycle to ensure sustainable nuclear energy utilization have been promoted. Decommissioning technologies for Tokai power station have been developed and accumulated know-how will be utilized in light water reactors. (T. Tanaka)

Science based integrated approach to advanced nuclear fuel development - integrated multi-scale multi-physics hierarchical modeling and simulation framework Part III: cladding

International Nuclear Information System (INIS)

Tome, Carlos N.; Caro, J.A.; Lebensohn, R.A.; Unal, Cetin; Arsenlis, A.; Marian, J.; Pasamehmetoglu, K.

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Reactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems to develop predictive tools is critical. Not only are fabrication and performance models needed to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. In this paper we review the current status of the advanced modeling and simulation of nuclear reactor cladding, with emphasis on what is available and what is to be developed in each scale of the project, how we propose to pass information from one scale to the next, and what experimental information is required for benchmarking and advancing the modeling at each scale level.

Materials Inventory Database for the Light Water Reactor Sustainability Program

Energy Technology Data Exchange (ETDEWEB)

Kazi Ahmed; Shannon M. Bragg-Sitton

2013-08-01

Scientific research involves the purchasing, processing, characterization, and fabrication of many sample materials. The history of such materials can become complicated over their lifetime – materials might be cut into pieces or moved to various storage locations, for example. A database with built-in functions to track these kinds of processes facilitates well-organized research. The Material Inventory Database Accounting System (MIDAS) is an easy-to-use tracking and reference system for such items. The Light Water Reactor Sustainability Program (LWRS), which seeks to advance the long-term reliability and productivity of existing nuclear reactors in the United States through multiple research pathways, proposed MIDAS as an efficient way to organize and track all items used in its research. The database software ensures traceability of all items used in research using built-in functions which can emulate actions on tracked items – fabrication, processing, splitting, and more – by performing operations on the data. MIDAS can recover and display the complete history of any item as a simple report. To ensure the database functions suitably for the organization of research, it was developed alongside a specific experiment to test accident tolerant nuclear fuel cladding under the LWRS Advanced Light Water Reactor Nuclear Fuels Pathway. MIDAS kept track of materials used in this experiment from receipt at the laboratory through all processes, test conduct and, ultimately, post-test analysis. By the end of this process, the database proved to be right tool for this program. The database software will help LWRS more efficiently conduct research experiments, from simple characterization tests to in-reactor experiments. Furthermore, MIDAS is a universal tool that any other research team could use to organize their material inventory.

Advanced light water reactor utility requirements document: Volume 1--ALWR policy and summary of top-tier requirements

International Nuclear Information System (INIS)

Anon.

1990-01-01

The U.S. utilities are leading an industry wide effort to establish the technical foundation for the design of the Advanced Light Water Reactor (ALWR). This effort, the ALWR Program, is being managed for the U.S. electric utility industry by the Electric Power Research Institute (EPRI) and includes participation and sponsorship of several international utility companies and close cooperation with the U.S. Department of Energy (DOE). The cornerstone of the ALWR Program is a set of utility design requirements which are contained in the ALWR Requirements Document. The purpose of the Requirement Document is to present a clear, complete statement of utility desires for their next generation of nuclear plants. The Requirements Document covers the entire plant up to the grid interface. It therefore is the basis for an integrated plant design, i.e., nuclear steam supply system and balance of plant, and it emphasizes those areas which are most important to the objective of achieving an ALWR which is excellent with respect to safety, performance, constructibility, and economics. The document applies to both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The Requirements Document is organized in three volumes. Volume 1 summarizes AlWR Program policy statements and top-tier requirements. The top-tier design requirements are categorized by major functions, including safety and investment protection, performance, and design process and constructibility. There is also a set of general design requirements, such as simplification and proven technology, which apply broadly to the ALWR design, and a set of economic goals for the ALWR program. The top-tier design requirements are described further in Volume 1 and are formally invoked as requirements in Volumes 2 and 3

Strategic Plan for Nuclear Energy -- Knowledge Base for Advanced Modeling and Simulation (NE-KAMS)

Energy Technology Data Exchange (ETDEWEB)

Rich Johnson; Kimberlyn C. Mousseau; Hyung Lee

2011-09-01

NE-KAMS knowledge base will assist computational analysts, physics model developers, experimentalists, nuclear reactor designers, and federal regulators by: (1) Establishing accepted standards, requirements and best practices for V&V and UQ of computational models and simulations, (2) Establishing accepted standards and procedures for qualifying and classifying experimental and numerical benchmark data, (3) Providing readily accessible databases for nuclear energy related experimental and numerical benchmark data that can be used in V&V assessments and computational methods development, (4) Providing a searchable knowledge base of information, documents and data on V&V and UQ, and (5) Providing web-enabled applications, tools and utilities for V&V and UQ activities, data assessment and processing, and information and data searches. From its inception, NE-KAMS will directly support nuclear energy research, development and demonstration programs within the U.S. Department of Energy (DOE), including the Consortium for Advanced Simulation of Light Water Reactors (CASL), the Nuclear Energy Advanced Modeling and Simulation (NEAMS), the Light Water Reactor Sustainability (LWRS), the Small Modular Reactors (SMR), and the Next Generation Nuclear Power Plant (NGNP) programs. These programs all involve computational modeling and simulation (M&S) of nuclear reactor systems, components and processes, and it is envisioned that NE-KAMS will help to coordinate and facilitate collaboration and sharing of resources and expertise for V&V and UQ across these programs. In addition, from the outset, NE-KAMS will support the use of computational M&S in the nuclear industry by developing guidelines and recommended practices aimed at quantifying the uncertainty and assessing the applicability of existing analysis models and methods. The NE-KAMS effort will initially focus on supporting the use of computational fluid dynamics (CFD) and thermal hydraulics (T/H) analysis for M&S of nuclear

Advanced light water reactor plants System 80+{trademark} design certification program. Annual progress report, October 1, 1994--September 30, 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-09-01

The purpose of this report is to provide the status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1995 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2, and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems.

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)

Development of next-generation light water reactor

International Nuclear Information System (INIS)

Ishibashi, Fumihiko; Yasuoka, Makoto

2010-01-01

The Next-Generation Light Water Reactor Development Program, a national project in Japan, was inaugurated in April 2008. The primary objective of this program is to meet the need for the replacement of existing nuclear power plants in Japan after 2030. With the aim of setting a global standard design, the reactor to be developed offers greatly improved safety, reliability, and economic efficiency through several innovative technologies, including a reactor core system with uranium enrichment of 5 to 10%, a seismic isolation system, long-life materials, advanced water chemistry, innovative construction techniques, optimized passive and active safety systems, innovative digital technologies, and so on. In the first three years, a plant design concept with these innovative features is to be established and the effectiveness of the program will be reevaluated. The major part of the program will be completed in 2015. Toshiba is actively engaged in both design studies and technology development as a founding member of this program. (author)

Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

International Nuclear Information System (INIS)

Noh, Hyung Gyun; Kang, Hie Chan

2016-01-01

In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins

Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

Energy Technology Data Exchange (ETDEWEB)

Noh, Hyung Gyun [Pohang University, Pohang (Korea, Republic of); Kang, Hie Chan [Kunsan University, Gunsan (Korea, Republic of)

2016-05-15

In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins.

Proceedings of the US Nuclear Regulatory Commission nineteenth water reactor safety information meeting

International Nuclear Information System (INIS)

Weiss, A.J.

1992-04-01

This three-volume report contains 83 papers out of the 108 that were presented at the Nineteenth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 28--30, 1991. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 14 different papers presented by researchers from Canada, Germany, France, Japan, Sweden, Taiwan, and USSR. This document, Volume 3, presents papers on: Structural engineering; Advanced reactor research; Advanced passive reactors; Human factors research; Human factors issues related to advanced passive light water researchers; Thermal Hydraulics; and Earth sciences. The individual papers have been cataloged separately

Review of light--water reactor safety studies. Volume 3 of health and safety impacts of nuclear, geothermal, and fossil-fuel electric generation in California

International Nuclear Information System (INIS)

Nero, A.V.; Farnaam, M.R.K.

1977-01-01

This report summarizes and compares important studies of light-water nuclear reactor safety, emphasizing the Nuclear Regulatory Commission's Reactor Safety Study, work on risk assessment funded by the Electric Power Research Institute, and the Report of the American Physical Society study group on light-water reactor safety. These reports treat risk assessment for nuclear power plants and provide an introduction to the basic issues in reactor safety and the needs of the reactor safety research program. Earlier studies are treated more briefly. The report includes comments on the Reactor Safety Study. The manner in which these studies may be used and alterations which would increase their utility are discussed

Study of the light and heavy water leaks in nuclear reactors and development of techniques for their detection, location and estimation

International Nuclear Information System (INIS)

Bashiruddin Mahmood, S.

1979-01-01

In heavy water type nuclear reactors the detection and control of heavy water and light water escapes from different systems is of vital importance in the successful and economic operation of these type of plants. The high cost of heavy water makes it imperative to minimise all such escapes, in order to reduce the loss as well as the upgrading cost of downgraded collection recovered from the reactor building. Original methods and devices have been developed at the Karachi Nuclear Power Plant which successfully solve this problem. This report describes the constructional and operational features of these devices

Advances in high temperature water chemistry and future issues

International Nuclear Information System (INIS)

Millett, P.J.

2005-01-01

This paper traces the development of advances in high temperature water chemistry with emphasis in the field of nuclear power. Many of the water chemistry technologies used in plants throughout the world today would not have been possible without the underlying scientific advances made in this field. In recent years, optimization of water chemistry has been accomplished by the availability of high temperature water chemistry codes such as MULTEQ. These tools have made the science of high temperature chemistry readily accessible for engineering purposes. The paper closes with a discussion of what additional scientific data and insights must be pursued in order to support the further development of water chemistry technologies for the nuclear industry. (orig.)

Passive Nuclear Plants Program (UPDATE)

International Nuclear Information System (INIS)

Chimeno, M. A.

1998-01-01

The light water passive plants program (PCNP), today Advanced Nuclear Power Plants Program (PCNA), was constituted in order to reach the goals of the Spanish Electrical Sector in the field of advanced nuclear power plants, optimize the efforts of all Spanish initiatives, and increase joint presence in international projects. The last update of this program, featured in revision 5th of the Program Report, reflects the consolidation of the Spanish sector's presence in International programs of the advanced power plants on the basis of the practically concluded American ALWR program. Since the beginning of the program , the PCNP relies on financing from the Electrical sector, Ocide, SEPI-Endesa, Westinghouse, General Electric, as well as from the industrial cooperators, Initec, UTE (Initec- Empresarios Agrupados), Ciemat, Enusa, Ensa and Tecnatom. The program is made up of the following projects, already concluded: - EPRI's Advanced Light Water Plants Certification Project - Westinghouse's AP600 Project - General Electric's SBWR Project (presently paralyzed) and ABWR project Currently, the following project are under development, at different degrees of advance: - EPP project (European Passive Plant) - EBWR project (European Advanced Boiling Water Reactor)

Testing Systems and Results for Advanced Nuclear Fuel Materials

International Nuclear Information System (INIS)

Rooyen, I.J. van; Griffith, G.W.; Garnier, J.E.

2012-01-01

Light Water Reactor Sustainability (LWRS) Program Advanced LWR Nuclear Fuel Development (ALFD) Pathway. Development and testing of high performance fuel cladding identified as high priority to support: enhancement of fuel performance, reliability, and reactor safety. One of the technologies being examined is an advanced fuel cladding made from ceramic matrix composites (CMC) utilizing silicon carbide (SiC) as a structural material supplementing a commercial Zircaloy-4 (Zr-4) tube. A series of out-of-pile tests to fully characterize the SiC CMC hybrid design to produce baseline data. The planned tests are intended to either produce quantitative data or to demonstrate the properties required to achieve two initial performance conditions relative to standard zircaloybased cladding: decreased hydrogen uptake (corrosion) and decreased fretting of the cladding tube under normal operating and postulated accident conditions. These two failure mechanisms account for approximately 70% of all in-pile failures of LWR commercial fuel assemblies

Water Treatment Using Advanced Ultraviolet Light Sources Final Report CRADA No. TC02089.0

Energy Technology Data Exchange (ETDEWEB)

Hoppes, W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Oster, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2017-08-15

This was a collaborative effort between Lawrence Livermore National Security, LLC as manager and operator of Lawrence Livermore National Laboratory (LLNL) and Teknichal Services, LLC (TkS), to develop water treatment systems using advanced ultraviolet light sources. The Russian institutes involved with this project were The High Current Electronics Institute (HCEI) and Russian Institute of Technical Physics-Institute of Experimental Physics (VNIIEF). HCEI and VNIIEF developed and demonstrated the potential commercial viability of short-wavelength ultraviolet excimer lamps under a Thrust 1 Initiatives for Proliferation Prevention (IPP) Program. The goals of this collaboration were to demonstrate both the commercial viability of excilampbased water disinfection and achieve further substantial operational improvement in the lamps themselves; particularly in the area of energy efficiency.

Nondestructive examination (NDE) reliability for inservice inspection of light water reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

1989-10-01

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvement that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements based on material properties, service conditions, and NDE capabilities and uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from October 1987 through March 1988. 21 refs., 28 figs., 2 tabs

Nondestructive examination (NDE) reliability for inservice inspection of light waters reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

1989-11-01

Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from April 1988 through September 1988. 33 refs., 70 figs., 12 tabs

Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Diaz, A.A.; Friley, J.R.

1993-09-01

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section XI of the ASME Code. This is a progress report covering the programmatic work from October 1991 through March 1992

Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

International Nuclear Information System (INIS)

Forsberg, C.W.

1989-01-01

Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' [Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety] is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document

The risks of nuclear energy technology. Safety concepts of light water reactors

Energy Technology Data Exchange (ETDEWEB)

Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Inst. fuer Kern- und Energietechnk (IKET); Kessler, Guenter; Veser, Anke; Schlueter, Franz-Hermann

2014-11-01

Analyses the risks of nuclear power stations. Discusses the security concept of reactors. Analyzes possible crash of air planes on a reactor containment. Presents measures against the spread of radioactivity after a severe accident. Written in engaging style for professionals and policy makers. The book analyses the risks of nuclear power stations. The security concept of reactors is explained. Measures against the spread of radioactivity after a severe accident, accidents of core melting and a possible crash of an air plane on a reactor containment are discussed. The book covers three scientific subjects of the safety concepts of Light Water Reactors: - A first part describes the basic safety design concepts of operating German Pressurized Water Reactors and Boiling Water Reactors including accident management measures introduced after the reactor accidents of Three Mile Island and Chernobyl. These safety concepts are also compared with the experiences of the Fukushima accidents. In addition, the safety design concepts of the future modern European Pressurized Water Reactor (EPR) and of the future modern Boiling Water Reactor SWR-1000 (KERENA) are presented. These are based on new safety research results of the past decades. - In a second, part the possible crash of military or heavy commercial air planes on a reactor containment is analyzed. It is shown that reactor containments can be designed to resist to such an airplane crash. - In a third part, an online decision system is presented. It allows to analyze the distribution of radioactivity in the atmosphere and to the environment after a severe reactor accident. It provides data for decisions to be taken by authorities for the minimization of radiobiological effects to the population. This book appeals to readers who have an interest in save living conditions and some understanding for physics or engineering.

A critical heat flux correlation for advanced pressurized light water reactor application

International Nuclear Information System (INIS)

Dalle Donne, M.; Hame, W.

1982-05-01

Many CHF-correlations have been developed for water cooled rod clusters representing typical PWR or BWR fuel element geometries with relative wide rod lattices. However the fuel elements of an Advanced Pressurized Water Reactor (APWR) have a tight fuel rod lattice, in view of increasing the fuel utilization. It was therefore decided to produce a new CHF-correlation valid for rod bundles with tight lattices. The already available WSC-2 correlation was chosen as a basis. The geometry dependent parameters of this correlation were determined again with the method of the root mean square fitting from the experimental data of the CHF-tests performed in the frame of the Light Water Breeder Reactor programme at the Bettis Laboratory. These tests include triangular array rod bundles with very tight lattices. Furthermore the effect of spiral spacer ribs was investigated on the basis of experimental data from the Columbia University. Application of the new CHF-correlation to conditions typical for an APWR shows that the predicted critical heat fluxes are much smaller than those calculated with the usual PWR-CHF-correlations, but they are higher than those predicted by the B+W-VPI+SU correlation. (orig.) [de

Impact of nuclear library difference on neutronic characteristics of thorium-loaded light water reactor fuel

International Nuclear Information System (INIS)

Unesaki, H.; Isaka, S.; Nakagome, Y.

2006-01-01

Impact of nuclear library difference on neutronic characteristics of thorium-loaded light water reactor fuel is investigated through cell burnup calculations using SRAC code system. Comparison of k ∞ and nuclide composition was made between the results obtained by JENDL-3.3, ENDF/B-VI.8 and JEFF3.0 for (U, Th)O 2 fuels as well as UO 2 fuels, with special interest on the burnup dependence of the neutronic characteristics. The impact of nuclear data library difference on k ∞ of (U, Th)O 2 fuels was found to be significantly large compared to that of UO 2 fuels. Notable difference was also found in nuclide concentration of TRU nuclides. (authors)

Handbook of advanced lighting technology

CERN Document Server

Sun, Ching-Cherng; Zissis, Georges; Ma, Ruiqing

2017-01-01

The Handbook of Advanced Lighting Technology is a major reference work on the subject of light source science and technology, with particular focus on solid-state light sources – LEDs and OLEDs – and the development of 'smart' or 'intelligent' lighting systems; and the integration of advanced light sources, sensors, and adaptive control architectures to provide tailored illumination which is 'fit to purpose.' The concept of smart lighting goes hand-in-hand with the development of solid-state light sources, which offer levels of control not previously available with conventional lighting systems. This has impact not only at the scale of the individual user, but also at an environmental and wider economic level. These advances have enabled and motivated significant research activity on the human factors of lighting, particularly related to the impact of lighting on healthcare and education, and the Handbook provides detailed reviews of work in these areas. The potential applications for smart lighting span ...

Status of advanced technology and design for water cooled reactors: Heavy water reactors

International Nuclear Information System (INIS)

1989-07-01

In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

Advances in heavy water reactors

International Nuclear Information System (INIS)

1994-03-01

The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The Technical Committee Meeting (TCM) on Advances in Heavy Water Reactors was organized by the IAEA in the framework of the activities of the International Working Group on Advanced Technologies for Water Cooled Reactors (IWGATWR) and hosted by the Atomic Energy of Canada Limited. Sixty-five participants from nine countries (Canada, Czech Republic, India, German, Japan, Republic of Korea, Pakistan, Romania and USA) and the IAEA attended the TCM. Thirty-four papers were presented and discussed in five sessions. A separate abstract was prepared for each of these papers. All recommendations which were addressed by the participants of the Technical Committee meeting to the IWGATWR have been submitted to the 5th IWGATWR meeting in September 1993. They were reviewed and used as input for the preparation of the IAEA programme in the area of advanced water cooled reactors. This TCM was mainly oriented towards advances in HWRs and on projects which are now in the design process and under discussion. Refs, figs and tabs

Fuel rod bundles proposed for advanced pressure tube nuclear reactors

International Nuclear Information System (INIS)

Prodea, Iosif; Catana, Alexandru

2010-01-01

The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACR TM -1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for 2013

Energy Technology Data Exchange (ETDEWEB)

Hallbert, Bruce [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Thomas, Ken [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

2014-09-01

Reliable instrumentation, information, and control (II&C) 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&C technology integration.

Thorium fuel for light water reactors - reducing proliferation potential of nuclear power fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Galperin, A; Radkowski, A [Ben-Gurion Univ. of the Negev, Beersheba (Israel)

1996-12-01

The proliferation potential of the light water reactor fuel cycle may be significantly reduced by utilization of thorium as a fertile component of the nuclear fuel. The main challenge of Th utilization is to design a core and a fuel cycle, which would be proliferation-resistant and economically feasible. This challenge is met by the Radkowsky Thorium Reactor (RTR) concept. So far the concept has been applied to a Russian design of a 1,000 MWe pressurized water reactor, known as a WWER-1000, and designated as VVERT. The following are the main results of the preliminary reference design: * The amount of Pu contained in the RTR spent fuel stockpile is reduced by 80% in comparison with a VVER of a current design. * The isotopic composition of the RTR-Pu greatly increases the probability of pre-initiation and yield degradation of a nuclear explosion. An extremely large Pu-238 content causes correspondingly large heat emission, which would complicate the design of an explosive device based on RTR-Pu. The economic incentive to reprocess and reuse the fissile component of the RTR spent fuel is decreased. The once-through cycle is economically optimal for the RTR core and cycle. To summarize all the items above: the replacement of a standard (U-based) fuel for nuclear reactors of current generation by the RTR fuel will provide an inherent barrier for nuclear weapon proliferation. This inherent barrier, in combination with existing safeguard measures and procedures is adequate to unambiguously disassociate civilian nuclear power from military nuclear power. * The RTR concept is applied to existing power plants to assure its economic feasibility. Reductions in waste disposal requirements, as well as in natural U and fabrication expenses, as compared to a standard WWER fuel, provide approximately 20% reduction in fuel cycle (authors).

Water chemistry features of advanced heavy water reactor

International Nuclear Information System (INIS)

Sriram, Jayasree; Vijayan, K.; Kain, Vivekanad; Velmurugan, S.

2015-01-01

Advanced Heavy Water Reactor (AHWR) being designed in India proposes to use Plutonium and Thorium as fuel. The objective is to extract energy from the uranium-233 formed from Thorium. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a natural circulation reactor. Thus, it has got several advanced passive safety features built into the system. The various water coolant systems are listed below. i) Main Heat transport System ii) Feed water system iii) Condenser cooling system iv) Process water system and safety systems. As it is a tube type reactor, the radiolysis control differs from the normal boiling water reactor. The coolant enters the bottom of the coolant channel, boiling takes place and then the entire steam water mixture exits the core through the long tail pipes and reaches the moisture separator. Thus, there is a need to devise methods to protect the tail pipes from oxidizing water chemistry condition. Similarly, the moderator heavy water coolant chemistry differs from that of moderator system chemistry of PHWR. The reactivity worth per ppm of gadolinium and boron are low in comparison to PHWR. As a result, much higher concentration of neutron poison has to be added for planned shutdown, start up and for actuating SDS-2. The addition of higher concentration of neutron poison result in higher radiolytic production of deuterium and oxygen. Their recombination back to heavy water has to take into account the higher production of these gases. This paper also discusses the chemistry features of safety systems of AHWR. In addition, the presentation will cover the chemistry monitoring methodology to be implemented in AHWR. (author)

[Design for constructability studies in support of the DOE ALWR (Advanced Light Water Reactor) Program

International Nuclear Information System (INIS)

1990-03-01

This technical report accounts for work performed as part of Duke Power Company's Design for Constructability Program. This program is contractual agreement AC03-86SF16566, part of the US Department of Energy's Technology Program in Support of Advanced Light Water Reactors. This report covers the period from contract inception (September 1986) through completion (March 1990). This report is divided into 4 volumes. Volume 1 includes the executive summary and significant program conclusions. The details supporting these conclusions are in Volume 3, Improving Construction Performance, and Volume 4, Enchancing Constructability Through Design. Volume 2 includes a description of the program, objectives, and approach. A significant conclusion from these discussions was the identification of a ''missing link'' in ALWR programs. With an essentially complete, certified design, the majority of the up-front planning and preparation for implementing the design can be accomplished. Though a monumental undertaking beyond the scope of this project, this up-front planning and preparation must be considered as the next logical step for standardization. Much of the planning can be repeated with future plants and marketed to recoup expenditures. Devoting resources to develop the standard design (evolutionary or passive) to a marketable, standard, and comprehensive plant package is essential to revitalizing the option of nuclear energy. The DOE should seriously consider devoting these resources as a logical extension of its ALWR support

Compatibility of the ultraviolet light-ozone system for laundry waste water treatment in nuclear power plants

International Nuclear Information System (INIS)

Matsuo, Toshiaki; Nishi, Takashi; Matsuda, Masami; Izumida, Tatsuo

1997-01-01

As an alternative treatment system for laundry waste water in nuclear power plants, a system was chosen in which such organic compounds as surfactant would be oxidized by ultraviolet (UV) light and ozone. The system compatibility, UV light source, and dissolved ozone concentration were examined through experiments. First, ozone gas was absorbed in the waste water. After the dissolved ozone concentration equilibrated at the desired value, the waste water was irradiated by a mercury lamp. Then, the time dependence of the concentrations of the organic compounds, the dissolved ozone, and the hydrogen peroxide were measured to estimate the treatment rate of the system. The mercury lamp with a 10 5 -Pa vapor pressure achieved large UV radiation and a treatment rate increase, leading to a compatible system without secondary waste generation. The effect of the dissolved ozone concentration on the treatment rate was saturated when concentration was >3.3 x 10 -4 mol/10 -3 m 3 at the time UV radiation was started. Numerical results indicated the saturation was due to hydrogen peroxide generation, which prevents hydroxyl radical generation

Nuclear calculation methods for light water moderated reactors

International Nuclear Information System (INIS)

Hicks, D.

1961-02-01

This report is intended as an introductory review. After a brief discussion of problems encountered in the nuclear design of water moderated reactors a comprehensive scheme of calculations is described. This scheme is based largely on theoretical methods and computer codes developed in the U.S.A. but some previously unreported developments made in this country are also described. It is shown that the effective reproduction factor of simple water moderated lattices may be estimated to an accuracy of approximately 1%. Methods for treating water gap flux peaking and control absorbers are presented in some detail, together with a brief discussion of temperature coefficients, void coefficients and burn-up problems. (author)

High Level Requirements for the Nuclear Energy -- Knowledge Base for Advanced Modeling and Simulation (NE-KAMS)

Energy Technology Data Exchange (ETDEWEB)

Rich Johnson; Hyung Lee; Kimberlyn C. Mousseau

2011-09-01

The US Department of Energy, Office of Nuclear Energy (DOE-NE), has been tasked with the important mission of ensuring that nuclear energy remains a compelling and viable energy source in the U.S. The motivations behind this mission include cost-effectively meeting the expected increases in the power needs of the country, reducing carbon emissions and reducing dependence on foreign energy sources. In the near term, to ensure that nuclear power remains a key element of U.S. energy strategy and portfolio, the DOE-NE will be working with the nuclear industry to support safe and efficient operations of existing nuclear power plants. In the long term, to meet the increasing energy needs of the U.S., the DOE-NE will be investing in research and development (R&D) and working in concert with the nuclear industry to build and deploy new, safer and more efficient nuclear power plants. The safe and efficient operations of existing nuclear power plants and designing, licensing and deploying new reactor designs, however, will require focused R&D programs as well as the extensive use and leveraging of advanced modeling and simulation (M&S). M&S will play a key role in ensuring safe and efficient operations of existing and new nuclear reactors. The DOE-NE has been actively developing and promoting the use of advanced M&S in reactor design and analysis through its R&D programs, e.g., the Nuclear Energy Advanced Modeling and Simulation (NEAMS) and Consortium for Advanced Simulation of Light Water Reactors (CASL) programs. Also, nuclear reactor vendors are already using CFD and CSM, for design, analysis, and licensing. However, these M&S tools cannot be used with confidence for nuclear reactor applications unless accompanied and supported by verification and validation (V&V) and uncertainty quantification (UQ) processes and procedures which provide quantitative measures of uncertainty for specific applications. The Nuclear Energy Knowledge base for Advanced Modeling and Simulation

New lineup of light water reactors

International Nuclear Information System (INIS)

Okamura, Kiyoshi; Oshima, Koichiro; Kitsukawa, Keisuke

2007-01-01

Toshiba is promoting technical studies for upcoming nuclear power plants based on its large accumulation of experience in boiling water reactor (BWR) design, manufacturing, construction, and maintenance. Our goal is to achieve higher reliability, lower life-cycle costs, and better competitiveness for nuclear power plants compared with other energy sources. In addition, we are developing a new light water reactor (LWR) lineup featuring the safest and most economical LWRs in the world as next-generation reactors almost at new construction and replacement in the Japanese and international markets expected to start from the 2020s. We are committed not only to developing BWRs with the world's highest performance but also to participating in the pressurized water reactor (PWR) market, taking advantage of the synergistic effect of both Toshiba's and Westinghouse's experience. (author)

Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Diaz, A.A.; Friley, J.R.; Good, M.S.; Greenwood, M.S.; Heasler, P.G.; Hockey, R.L.; Kurtz, R.J.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

1992-07-01

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWR's); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section XI of the ASME Code. This is a progress report covering the programmatic work from April 1991 through September 1991

Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Diaz, A.A.; Friley, J.R.; Greenwood, M.S.; Heasler, P.G.; Kurtz, R.J.; Simonen, F.A.; Spanner, J.C.; Vo, T.V.

1993-11-01

The Evaluation and Improvement of NDE Reliability for Inservice inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs);using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel and other components inspected in accordance with Section XI of the ASME Code. This is a programs report covering the programmatic work from April 1992 through September 1992

Nondestructive examination (NDE) reliability for inservice inspection of light water reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

1991-08-01

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section 6 of the ASME Code. This is a progress report covering the pro grammatic work from April 1989 through September 1989. 12 refs., 4 figs. 5 tabs

Nuclear desalination in the Arab world - Part II: Advanced inherent and passive safe nuclear reactors

International Nuclear Information System (INIS)

Karameldin, A.; Samer S. Mekhemar

2004-01-01

Rapid increases in population levels have led to greater demands for fresh water and electricity in the Arab World. Different types of energies are needed to contribute to bridging the gap between increased demand and production. Increased levels of safeguards in nuclear power plants have became reliable due to their large operational experience, which now exceeds 11,000 years of operation. Thus, the nuclear power industry should be attracting greater attention. World electricity production from nuclear power has risen from 1.7% in 1970 to 17%-20% today. This ratio had increased in June 2002 to reach more than 30%, 33% and 42% in Europe, Japan, and South Korea respectively. In the Arab World, both the public acceptance and economic viability of nuclear power as a major source of energy are greatly dependent on the achievement of a high level of safety and environmental protection. An assessment of the recent generation of advanced reactor safety criteria requirements has been carried out. The promising reactor designs adapted for the Arab world and other similar developing countries are those that profit from the enhanced and passive safety features of the new generation of reactors, with a stronger focus on the effective use of intrinsic characteristics, simplified plant design, and easy construction, operation and maintenance. In addition, selected advanced reactors with a full spectrum from small to large capacities, and from evolutionary to radical types, which have inherent and passive safety features, are discussed. The relevant economic assessment of these reactors adapted for water/electricity cogeneration have been carried out and compared with non-nuclear desalination methods. This assessment indicates that, water/electricity cogeneration by the nuclear method with advanced inherent and passive safe nuclear power plants, is viable and competitive. (author)

Advanced marine reactor MRX and application to nuclear barge supplying electricity and heat

International Nuclear Information System (INIS)

Ishida, Toshihisa; Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Fukuhara, Yoshifumi; Ochiai, Masa-aki

2000-01-01

The basic design concept of an advanced marine reactor MRX has been established with adoption of several new technologies. The MRX is an integral-type PWR with 100 MWt aimed basically for use of ship propulsion. Adoption of a water-filled containment together with the integral type reactor makes the reactor light-weight and compact greatly. A engineered safety system is a simplified passive system, function of which is confirmed by the safety analysis. The MRX can be applied to an energy supply system of electricity and heat co-generation by installing it on a barge. Concept of a nuclear barge with the MRX of 334 MWt output is presented for use of supplying electricity, fresh water and hot water. Combined system of electric generation and desalination with the RO process can deliver variable output of electricity and fresh water according a demand. Latent heat of the exhausted steam from the turbine can be used effectively to raise the temperature of cold water as heat supply. (author)

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

International Nuclear Information System (INIS)

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

1981-09-01

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

The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

International Nuclear Information System (INIS)

Bruschi, H.J.

1996-01-01

At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse's advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

Energy Technology Data Exchange (ETDEWEB)

Bruschi, H.J. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1996-10-01

At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse`s advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

Calculation methods for advanced concept light water reactor lattices

International Nuclear Information System (INIS)

Carmona, S.

1986-01-01

In the last few years s several advanced concepts for fuel rod lattices have been studied. Improved fuel utilization is one of the major aims in the development of new fuel rod designs and lattice modifications. By these changes s better performance in fuel economics s fuel burnup and material endurance can be achieved in the frame of the well-known basic Light Water Reactor technology. Among the new concepts involved in these studies that have attracted serious attention are lattices consisting of arrays of annular rods duplex pellet rods or tight multicells. These new designs of fuel rods and lattices present several computational problems. The treatment of resonance shielded cross sections is a crucial point in the analyses of these advanced concepts . The purpose of this study was to assess adequate approximation methods for calculating as accurately as possible, resonance shielding for these new lattices. Although detailed and exact computational methods for the evaluation of the resonance shielding in these lattices are possible, they are quite inefficient when used in lattice codes. The computer time and memory required for this kind of computations are too large to be used in an acceptable routine manner. In order to over- come these limitations and to make the analyses possible with reasonable use of computer resources s approximation methods are necessary. Usual approximation methods, for the resonance energy regions used in routine lattice computer codes, can not adequately handle the evaluation of these new fuel rod lattices. The main contribution of the present work to advanced lattice concepts is the development of an equivalence principle for the calculation of resonance shielding in the annular fuel pellet zone of duplex pellets; the duplex pellet in this treatment consists of two fuel zones with the same absorber isotope in both regions. In the transition from a single duplex rod to an infinite array of this kind of fuel rods, the similarity of the

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

International Nuclear Information System (INIS)

Monteleone, S.

1994-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-04-01

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

Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for FY 2016

Energy Technology Data Exchange (ETDEWEB)

Hallbert, Bruce Perry [Idaho National Lab. (INL), Idaho Falls, ID (United States); Thomas, Kenneth David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-10-01

Reliable instrumentation, information, and control (II&C) 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&C technology integration.

77 FR 73056 - Initial Test Programs for Water-Cooled Nuclear Power Plants

Science.gov (United States)

2012-12-07

... Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for comment... (DG), DG-1259, ``Initial Test Programs for Water-Cooled Nuclear Power Plants.'' This guide describes... (ITPs) for light water cooled nuclear power plants. DATES: Submit comments by January 31, 2013. Comments...

Advanced Light Source (ALS)

Data.gov (United States)

Federal Laboratory Consortium — The Advanced Light Source (ALS), a world leader in soft x-ray science, generates light in the wavelengths needed for examining the atomic and electronic structure of...

LBB application in the US operating and advanced reactors

Energy Technology Data Exchange (ETDEWEB)

Wichman, K.; Tsao, J.; Mayfield, M.

1997-04-01

The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

Terms for describing new, advanced nuclear power plants

International Nuclear Information System (INIS)

1997-04-01

The IAEA's Division of Nuclear Power and the Fuel Cycle (then the Division of Nuclear Power) took an initiative in this field some years ago when work was initiated in the area of ''safety related terms'' by its International Working Group on Advanced Technologies for Water Cooled Reactors. This activity drew on advice from reactor design organizations, research institutes and government organizations, and aimed at helping eliminate confusion and misuse of safety related terms in widespread use, clarifying technical thinking regarding these terms, and improving nuclear power acceptability by providing precisely described technical meanings to them. After discussion also in the International Working Groups for Gas Cooled Reactors and Fast Reactors, the work resulted in the publication in September 1991 of IAEA-TECDOC-626, entitled ''Safety Related Terms for Advanced Nuclear Plants'', which has become a widely used publication. The present TECDOC has been prepared using the same approach to obtain advice from involved parties. Drafts of this report have been reviewed by the International Working Groups on Water Cooled Reactors, Fast Reactors and Gas Cooled Reactors, as well as by the IAEA's International Fusion Research Council (IFRC). The comments and suggestions received have been evaluated and utilized for producing the present TECDOC. 3 figs

The Westinghouse AP600 an advanced nuclear option for small or medium electricity grids

International Nuclear Information System (INIS)

Bruschi, H. J.; Novak, V.

1996-01-01

During the early days of commercial nuclear power, many countries looking to add nuclear power to their energy mix required large plants to meet the energy needs of rapidly growing populations and large industrial complexes. The majority of plants worldwide are in the range of 100 megawatts and beyond. During the 1970s, it became apparent that a smaller nuclear plants would appeal to utilities looking to add additional power capacity to existing grids, or to utilities in smaller countries which were seeking efficient, new nuclear generation capacity for the first time. For instance, the Westinghouse-designed 600 megawatt Krsko plant in Slovenia began operation in 1980, providing electricity to inhabitants of relatively small, yet industrial populations of Slovenia and Croatia. This plant design incorporated the best, proven technology available at that time, based on 20 years of Westinghouse PWR pioneering experience. Beginning in the early 1980s, Westinghouse began to build further upon that experience - in part through the advanced light water reactor programs established by the Electric Power Research institute (EPRI) and the U.S. Department of Energy (DOE) - to design a simplified, advanced nuclear reactor in the 600 megawatt range. Originally, Westinghouse's development of its AP600 (advanced, passive 600-megawatt) plants was geared towards the needs of U.S. utilities which specified smaller, simplified nuclear options for the decades ahead. It soon became evident that the small and medium sized electricity grids of international markets could benefit from this new reactor. From the earliest days of Westinghouse's AP600 development, the corporation invited members of the international nuclear community to take part in the design, development and testing of the AP600 - with the goal of designing a reactor that would meet the diverse needs of an international industry composed of countries with similar, yet different, concerns. (author)

Proceedings of the 2006 international congress on advances in nuclear power plants - ICAPP'06

International Nuclear Information System (INIS)

2006-01-01

Following the highly successful ICAPP'05 meeting held in Seoul Korea, the 2006 International Congress on Advances in Nuclear Power Plants brought together international experts of the nuclear industry involved in the operation, development, building, regulation and research related to Nuclear Power Plants. The program covers the full spectrum of Nuclear Power Plant issues from design, deployment and construction of plants to research and development of future designs and advanced systems. The program covers lessons learned from power, research and demonstration reactors from over 50 years of experience with operation and maintenance, structures, materials, technical specifications, human factors, system design and reliability. The program by technical track deals with: - 1. Water-Cooled Reactor Programs and Issues Evolutionary designs, innovative, passive, light and heavy water cooled reactors; issues related to meeting medium term utility needs; design and regulatory issues; business, political and economic challenges; infrastructure limitations and improved construction techniques including modularization. - 2. High Temperature Gas Cooled Reactors Design and development issues, components and materials, safety, reliability, economics, demonstration plants and environmental issues, fuel design and reliability, power conversion technology, hydrogen production and other industrial uses; advanced thermal and fast reactors. - 3. Long Term Reactor Programs and Strategies Reactor technology with enhanced fuel cycle features for improved resource utilization, waste characteristics, and power conversion capabilities. Potential reactor designs with longer development times such as, super critical water reactors, liquid metal reactors, gaseous and liquid fuel reactors, Gen IV, INPRO, EUR and other programs. - 4. Operation, Performance and Reliability Management Training, O and M costs, life cycle management, risk based maintenance, operational experiences, performance and

A new book : 'light-water reactor materials'

International Nuclear Information System (INIS)

Olander, Donald R.; Motta, Arthur T.

2005-01-01

The contents of a new book currently in preparation are described. The dearth of books in the field of nuclear materials has left both students in nuclear materials classes and professionals in the same field without a resource for the broad fundamentals of this important sub-discipline of nuclear engineering. The new book is devoted entirely to materials problems in the core of light-water reactors, from the pressure vessel into the fuel. Key topics deal with the UO 2 fuel, zircaloy cladding, stainless steel, and of course, water. The restriction to LWR materials does not mean a short monograph; the enormous quantity of experimental and theoretical work over the past 50 years on these materials presents a challenge of culling the most important features and explaining them in the simplest quantitative fashion. Moreover, LWRs will probably be the sole instrument of the return of nuclear energy in electric power production for the next decade or so. By that time, a new book will be needed

Analysis of an accelerator-driven subcritical light water reactor

International Nuclear Information System (INIS)

Kruijf, W.J.M. de; Wakker, P.H.; Wetering, T.F.H. van de; Verkooijen, A.H.M.

1997-01-01

An analysis of the basic characteristics of an accelerator-driven light water reactor has been made. The waste in the nuclear fuel cycle is considerably less than in the light water reactor open fuel cycle. This is mainly caused by the use of equilibrium nuclear fuel in the reactor. The accelerator enables the use of a fuel composition with infinite multiplication factor k ∞ < 1. The main problem of the use of this type of fuel is the strongly peaked flux distribution in the reactor core. A simple analytical model shows that a large core is needed with a high peak power factor in order to generate net electric energy. The fuel in the outer regions of the reactor core is used very poorly. 7 refs., 4 figs., 1 tab

Design features of the Light Water Breeder Reactor (LWBR) which improve fuel utilization in light water reactors (LWBR development program)

International Nuclear Information System (INIS)

Hecker, H.C.; Freeman, L.B.

1981-08-01

This report surveys reactor core design features of the Light Water Breeder Reactor which make possible improved fuel utilization in light water reactor systems and breeding with the uranium-thorium fuel cycle. The impact of developing the uranium-thorium fuel cycle on utilization of nuclear fuel resources is discussed. The specific core design features related to improved fuel utilization and breeding which have been implemented in the Shippingport LWBR core are presented. These design features include a seed-blanket module with movable fuel for reactivity control, radial and axial reflcetor regions, low hafnium Zircaloy for fuel element cladding and structurals, and a closely spaced fuel rod lattice. Also included is a discussion of several design modifications which could further improve fuel utilization in future light water reactor systems. These include further development of movable fuel control, use of Zircaloy fuel rod support grids, and fuel element design modifications

Proceedings of the 2008 International Congress on Advances in Nuclear Power Plants - ICAPP '08

International Nuclear Information System (INIS)

2008-01-01

ICAPP 2008 congress brought together international experts of the nuclear industry involved in the operation, development, building, regulation and research related to Nuclear Power Plants. The program covered the full spectrum of Nuclear Power Plant issues from design, deployment and construction of plants to research and development of future designs and advanced systems. It covered also lessons learned from power, research and demonstration reactors from over 50 years of experience with operation and maintenance, structures, materials, technical specifications, human factors, system design and reliability. The program comprised 13 technical tracks: 1. Water-Cooled Reactor Programs and Issues: Evolutionary designs, innovative, passive, light and heavy water cooled reactors; issues related to meeting near term utility needs; design issues; business, economical cost challenges; infrastructure limitations and improved construction techniques including modularization. 2. High Temperature Gas Cooled Reactors: Design and development issues, components and materials, safety, reliability, economics, demonstration plants and environmental issues, fuel design and reliability, power conversion technology, impact of non electricity applications on reactor design; advanced thermal and fast reactors. 3. LMFR and Longer Term Reactor Programs: Reactor technology with enhanced fuel cycle features for improved resource utilization, waste characteristics, and power conversion capabilities. Potential reactor designs with longer development times such as super critical water reactors and liquid fuel reactors, Gen IV, INPRO, EUR and other programs. 4. Operation, Performance and Reliability Management: Training, O and M costs, life cycle management, risk based maintenance, operational experiences, performance and reliability improvements, outage optimization, human factors, plant staffing, outage reduction features, major component reliability, repair and replacement, in

Advanced nuclear plant design options to cope with external events

International Nuclear Information System (INIS)

2006-02-01

With the stagnation period of nuclear power apparently coming to an end, there is a renewed interest in many Member States in the development and application of nuclear power plants (NPPs) with advanced reactors. Decisions on the construction of several NPPs with evolutionary light water reactors have been made (e.g. EPR Finland for Finland and France) and more are under consideration. There is a noticeable progress in the development and demonstration of innovative high temperature gas cooled reactors, for example, in China, South Africa and Japan. The Generation IV International Forum has defined the International Near Term Deployment programme and, for a more distant perspective, six innovative nuclear energy systems have been selected and certain R and D started by several participating countries. National efforts on design and technology development for NPPs with advanced reactors, both evolutionary and innovative, are ongoing in many Member States. Advanced NPPs have an opportunity to be built at many sites around the world, with very broad siting conditions. There are special concerns that safety of these advanced reactors may be challenged by external events following new scenarios and failure modes, different from those well known for the currently operated reactors. Therefore, the engineering community identified the need to assess the proposed design configurations in relation to external scenarios at the earliest stages of the design development. It appears that an early design optimization in relation to external events is a necessary requirement to achieve safe and economical advanced nuclear power plants. Reflecting on these developments, the IAEA has planned the preparation of a report to define design options for protection from external event impacts in NPPs with evolutionary and innovative reactors. The objective of this publication is to present the state-of-the-art in design approaches for the protection of NPPs with evolutionary and innovative

Advanced digital technology - improving nuclear power plant performance through maintainability

International Nuclear Information System (INIS)

Ford, J.L.; Senechal, R.R.; Altenhein, G.D.; Harvey, R.P.

1998-01-01

In today's energy sector there is ever increasing pressure on utilities to operate power plants at high capacity factors. To ensure nuclear power is competitive into the next century, it is imperative that strategic design improvements be made to enhance the performance of nuclear power plants. There are a number of factors that affect a nuclear power plant's performance; lifetime maintenance is one of the major contributors. The maturing of digital technology has afforded ABB the opportunity to make significant design improvements in the area of maintainability. In keeping with ABB's evolutionary advanced nuclear plant design approach, digital technology has systematically been incorporated into the control and protection systems of the most recent Korean nuclear units in operation and under construction. One example of this was the multi-functional design team approach that was utilized for the development of ABB's Digital Plant Protection System (DPPS). The design team consisted of engineers, maintenance technicians, procurement specialists and manufacturing personnel in order to provide a complete perspective on all facets of the design. The governing design goals of increased reliability and safety, simplicity of design, use of off-the-shelf products and reduced need for periodic surveillance testing were met with the selection of proven ABB-Advant Programmable Logic Controllers (PLCs) as the heart of the DPPS. The application of digital PLC technology allows operation for extended periods without requiring routine maintenance or re-calibration. A well documented commercial dedication program approved by the United States Nuclear Regulatory Commission (US NRC) as part of the System 80+ TM Advanced Light Water Reactor Design Certification Program, allowed the use of off-the shelf products in the design of the safety protection system. In addition, a number of mechanical and electrical improvements were made which support maintainability. The result is a DPPS

Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

International Nuclear Information System (INIS)

1989-02-01

Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

Cooling water requirements and nuclear power plants

International Nuclear Information System (INIS)

Rao, T.S.

2010-01-01

Indian nuclear power programme is poised to scuttle the energy crisis of our time by proposing joint ventures for large power plants. Large fossil/nuclear power plants (NPPs) rely upon water for cooling and are therefore located near coastal areas. The amount of water a power station uses and consumes depends on the cooling technology used. Depending on the cooling technology utilized, per megawatt existing NPPs use and consume more water (by a factor of 1.25) than power stations using other fuel sources. In this context the distinction between 'use' and 'consume' of water is important. All power stations do consume some of the water they use; this is generally lost as evaporation. Cooling systems are basically of two types; Closed cycle and Once-through, of the two systems, the closed cycle uses about 2-3% of the water volumes used by the once-through system. Generally, water used for power plant cooling is chemically altered for purposes of extending the useful life of equipment and to ensure efficient operation. The used chemicals effluent will be added to the cooling water discharge. Thus water quality impacts on power plants vary significantly, from one electricity generating technology to another. In light of massive expansion of nuclear power programme there is a need to develop new ecofriendly cooling water technologies. Seawater cooling towers (SCT) could be a viable option for power plants. SCTs can be utilized with the proper selection of materials, coatings and can achieve long service life. Among the concerns raised about the development of a nuclear power industry, the amount of water consumed by nuclear power plants compared with other power stations is of relevance in light of the warming surface seawater temperatures. A 1000 MW power plant uses per day ∼800 ML/MW in once through cooling system; while SCT use 27 ML/MW. With the advent of new marine materials and concrete compositions SCT can be constructed for efficient operation. However, the

A contingency safe, responsible, economic, increased capacity spent nuclear fuel (SNF) advance fuel cycle

International Nuclear Information System (INIS)

Levy, S.

2008-01-01

The purpose of this paper is to have an Advanced Light Water (LWR) fuel cycle and an associated development program to provide a contingency plan to the current DOE effort to license once-through spent Light Water Reactor (LWR) fuel for disposition at Yucca Mountain (YM). The intent is to fully support the forthcoming June 2008 DOE submittal to the Nuclear Regulatory Commission (NRC) based upon the latest DOE draft DOE/EIS-0250F-SID dated October 2007 which shows that the latest DOE YM doses would readily satisfy the anticipated NRC and Environmental Protection Agency (EP) standards. The proposed Advance Fuel Cycle can offer potential resolution of obstacles that might arise during the NRC review and, particularly, during the final hearings process to be held in Nevada. Another reason for the proposed concept is that a substantial capacity growth of the YM repository will be necessary to accommodate the SNF of Advance Light Water Reactors (ALWRs) currently under consideration for United States (U.S.) electricity production (1) and the results of the recently issued study by the Electric Power Research Institute (EPRI) to reduce CO 2 emissions (2). That study predicts that by 2030 U.S. nuclear power generation would grow by 64 Gigawatt electrical (GWe) and account for 25.5 percent of the overall U.S. electrical generation. The current annual SNF once-through fuel cycle accumulation would rise from 2000-2100 MT (Metric Tons) to about 3480 MT in 2030 and the total SNF inventory, would reach nearly 500,000 MT by 2100 if U. S. nuclear power continues to grow at 1.1 percent per year after 2030. That last projection does not account for any SNF reduction due to increased fuel burnup or any increased capacity needed 'to establish supply Global Nuclear Energy Partnership (GNEP,) arrangements among nations to provide nuclear fuel and taking back spent fuel for recycling without spreading enrichment and reprocessing technologies' (3). The anticipated capacity of 120 MT

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

Proceedings of the 2004 international congress on advances in nuclear power plants - ICAPP'04

International Nuclear Information System (INIS)

2004-01-01

The 2004 International Congress on Advances in Nuclear Power Plants (ICAPP'04) provides a forum for the industry to exchange the latest ideas and research findings on nuclear plants from all perspectives. This conference builds on the success of last year's meeting held in Cordoba, Spain, and on the 2002 inaugural meeting held in Hollywood, Florida. Because of the hard work of many volunteers from around the world, ICAPP'04 has been successful in achieving its goal. More than 325 invited and contributed papers/presentations are part of this ICAPP. There are 5 invited plenary sessions and 70 technical sessions with contributed papers. The ICAPP'04 Proceedings contain almost 275 papers prepared by authors from 25 countries covering topics related to advances in nuclear power plant technology. The program by technical track deals with: 1 - Water-Cooled Reactor Programs and Issues (Status of All New Water-Cooled Reactor Programs; Advanced PWRs: Developmental Stage I; Advanced PWRs: Developmental Stage II; Advanced PWRs: Basic Design Stage; Advanced BWRs; Economics, Regulation, Licensing, and Construction; AP1000); 2 - High Temperature Gas Cooled Reactors (Pebble Bed Modular Reactors; Very High Temperature Reactors; HTR Fuels and Materials; Innovative HTRs and Fuel Cycles); 3 - Long Term Reactor Programs and Strategies (Supercritical Pressure Water Reactors; Lead-Alloy Fast Reactors; Sodium and Gas Fast Reactors; Status of Advanced Reactor Programs; Non-classical Reactor Concepts); 4 - Operation, Performance, and Reliability Management (Information Technology Effect on Plant Operation; Operation, Maintenance and Reliability; Improving Performance and Reducing O and M Costs; Plant Modernization and Retrofits); 5 - Plant Safety Assessment and Regulatory Issues (LOCA and non-LOCA Analysis Methodologies; LOCA and non-LOCA Plant Analyses; In-Vessel Retention; Containment Performance and Hydrogen Control; Advances in Severe Accident Analysis; Advances in Severe Accident

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.

Advanced lighting guidelines: 1993. Final report

Energy Technology Data Exchange (ETDEWEB)

Eley, C.; Tolen, T.M. [Eley Associates, San Francisco, CA (United States); Benya, J.R. [Luminae Souter Lighting Design, San Francisco, CA (United States); Rubinstein, F.; Verderber, R. [Lawrence Berkeley Lab., CA (United States)

1993-12-31

The 1993 Advanced Lighting Guidelines document consists of twelve guidelines that provide an overview of specific lighting technologies and design application techniques utilizing energy-efficient lighting practice. Lighting Design Practice assesses energy-efficient lighting strategies, discusses lighting issues, and explains how to obtain quality lighting design and consulting services. Luminaires and Lighting Systems surveys luminaire equipment designed to take advantage of advanced technology lamp products and includes performance tables that allow for accurate estimation of luminaire light output and power input. The additional ten guidelines -- Computer-Aided Lighting Design, Energy-Efficient Fluorescent Ballasts, Full-Size Fluorescent Lamps, Compact Fluorescent Lamps, Tungsten-Halogen Lamps, Metal Halide and HPS Lamps, Daylighting and Lumen Maintenance, Occupant Sensors, Time Scheduling Systems, and Retrofit Control Technologies -- each provide a product technology overview, discuss current products on the lighting equipment market, and provide application techniques. This document is intended for use by electric utility personnel involved in lighting programs, lighting designers, electrical engineers, architects, lighting manufacturers` representatives, and other lighting professionals.

The United States Advanced Reactor Technologies Research and Development Program

International Nuclear Information System (INIS)

O’Connor, Thomas J.

2014-01-01

The following aspects are addressed: • Nuclear energy mission; • Reactor research development and deployment (RD&D) programs: - Light Water Reactor Sustainability Program; - Small Modular Reactor Licensing Technical Support; - Advanced Reactor Technologies (ART)

A case study for INPRO methodology based on Indian advanced heavy water reactor

International Nuclear Information System (INIS)

Anantharaman, K.; Saha, D.; Sinha, R.K.

2004-01-01

Under Phase 1A of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) a methodology (INPRO methodology) has been developed which can be used to evaluate a given energy system or a component of such a system on a national and/or global basis. The INPRO study can be used for assessing the potential of the innovative reactor in terms of economics, sustainability and environment, safety, waste management, proliferation resistance and cross cutting issues. India, a participant in INPRO program, is engaged in a case study applying INPRO methodology based on Advanced Heavy Water Reactor (AHWR). AHWR is a 300 MWe, boiling light water cooled, heavy water moderated and vertical pressure tube type reactor. Thorium utilization is very essential for Indian nuclear power program considering the indigenous resource availability. The AHWR is designed to produce most of its power from thorium, aided by a small input of plutonium-based fuel. The features of AHWR are described in the paper. The case study covers the fuel cycle, to be followed in the near future, for AHWR. The paper deals with initial observations of the case study with regard to fuel cycle issues. (authors)

Indian advanced nuclear reactors

International Nuclear Information System (INIS)

Saha, D.; Sinha, R.K.

2005-01-01

For sustainable development of nuclear energy, a number of important issues like safety, waste management, economics etc. are to be addressed. To do this, a number of advanced reactor designs as well as fuel cycle technologies are being pursued worldwide. The advanced reactors being developed in India are the AHWR and the CHTR. Both the reactors use thorium based fuel and have many passive features. This paper describes the Indian advanced reactors and gives a brief account of the international initiatives for the sustainable development of nuclear energy. (author)

New or improved computational methods and advanced reactor design

International Nuclear Information System (INIS)

Nakagawa, Masayuki; Takeda, Toshikazu; Ushio, Tadashi

1997-01-01

Nuclear computational method has been studied continuously up to date, as a fundamental technology supporting the nuclear development. At present, research on computational method according to new theory and the calculating method thought to be difficult to practise are also continued actively to find new development due to splendid improvement of features of computer. In Japan, many light water type reactors are now in operations, new computational methods are induced for nuclear design, and a lot of efforts are concentrated for intending to more improvement of economics and safety. In this paper, some new research results on the nuclear computational methods and their application to nuclear design of the reactor were described for introducing recent trend of the nuclear design of the reactor. 1) Advancement of the computational method, 2) Reactor core design and management of the light water reactor, and 3) Nuclear design of the fast reactor. (G.K.)

Surveillance tests for light-water cooled nuclear power reactor vessels in IMEF

International Nuclear Information System (INIS)

Choo, Yong-Sun; Ahn, Sang-Bok; Park, Dae-Gyu; Jung, Yang-Hong; Yoo, Byung-Ok; Oh, Wan-Ho; Baik, Seung-Je; Koo, Dae-Seo; Lee, Key-Soon

1999-01-01

The surveillance tests for light-water cooled nuclear power reactor vessels were established to monitor the radiation-induced changes in the mechanical properties of ferritic materials in the beltline according to US NRC 10 CFR 50 App. G, US NRC RG1.99-rev.2, ASTM E185-82 and E185-94 in Irradiated Materials Examination Facility(IMEF). The surveillance capsule was transported from NPPs pool sites to KAERI IMEF by using a shipping cask. The capsule was cut and dismantled by capsule cutting machine and milling machine in M2 hot cell. Charpy tests and tension tests were performed in M5a and M5b hot cells respectively. Especially the EPMA located at hot lab was used to analyze the Ni and Cu wt% composition of base metal and weld for predicting the adjusted reference temperature(ART). The established process and test results were summarized in this paper. (author)

Framatome advanced nuclear power-benefits for our clients from the new company

International Nuclear Information System (INIS)

Weber, P.

2001-01-01

Framatome ANP (Advanced Nuclear Power) merges the complementary strengths of two global nuclear industry leaders Framatome and Siemens - offering clients the best technological solutions for safe, reliable and economical plant performance. With a combined workforce of 13,300 skilled individuals, Framatome ANP is now the nuclear industry's leading supplier. Serving as Original Equipment Manufacturer (OEM) for more than 90 reactors that provide about 30% of the world's total installed nuclear power capacity, our experienced resources remain focused on the local needs of individual clients, wherever in the world they may be. The Company main business used to be turnkey construction of complete Nuclear Power plants, BWR and PWR capabilities, heavy equipment manufacturing, comprehensive I and C capabilities, and also expertise and knowledge of VVER. Framatome ANP will benefit in all of its fields of activity of the experience gained through Framatome and Siemens' collaboration on the next generation reactor, the EPR, as well as on steam generators replacements and or modernization of VVER. Framatome ANP nuclear fuel designs for both PWR and BWR plants provide innovative features and world-leading performance. Framatome ANP is organized according a matrix organization with: - 4 Business Groups (Project and Engineering, Service, Nuclear Fuel, Mechanical Equipment) - 3 Regional Divisions (Framatome Advanced Nuclear Power S.A.S., France; Framatome Advanced Nuclear Power GmbH, Germany; Framatome Advanced Nuclear Power Inc., USA). By 30th January 2001 Siemens Nuclear Power GmbH, founded in 2000 as successor of the Nuclear Division of Siemens Power Generation Group (KWU), was renamed to Framatome Advanced Nuclear Power GmbH forming the German part of the world wide acting company. Over the past 40 years 23 nuclear power plants all around the world - not only pressurized and boiling water reactors, but also two heavy-watermoderated reactors have been designed, constructed and

Performance of advanced self-shielding models in DRAGON Version4 on analysis of a high conversion light water reactor lattice

International Nuclear Information System (INIS)

Karthikeyan, Ramamoorthy; Hebert, Alain

2008-01-01

A high conversion light water reactor lattice has been analysed using the code DRAGON Version4. This analysis was performed to test the performance of the advanced self-shielding models incorporated in DRAGON Version4. The self-shielding models are broadly classified into two groups - 'equivalence in dilution' and 'subgroup approach'. Under the 'equivalence in dilution' approach we have analysed the generalized Stamm'ler model with and without Nordheim model and Riemann integration. These models have been analysed also using the Livolant-Jeanpierre normalization. Under the 'subgroup approach', we have analysed Statistical self-shielding model based on physical probability tables and Ribon extended self-shielding model based on mathematical probability tables. This analysis will help in understanding the performance of advanced self-shielding models for a lattice that is tight and has a large fraction of fissions happening in the resonance region. The nuclear data for the analysis was generated in-house. NJOY99.90 was used for generating libraries in DRAGLIB format for analysis using DRAGON and A Compact ENDF libraries for analysis using MCNP5. The evaluated datafiles were chosen based on the recommendations of the IAEA Co-ordinated Research Project on the WIMS Library Update Project. The reference solution for the problem was obtained using Monte Carlo code MCNP5. It was found that the Ribon extended self-shielding model based on mathematical probability tables using correlation model performed better than all other models

Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

CERN Document Server

American Society for Testing and Materials. Philadelphia

2003-01-01

1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

International Nuclear Information System (INIS)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

The management plan of liquid effluent in Korean advanced light water reactor

International Nuclear Information System (INIS)

Kim, S. H.; Lim, H. S.; Jeong, D. W.; Jeong, D. Y.

2001-01-01

Non-radioactive liquid effluent in Korean Advanced Light Water Reactor is transferred and treated in centralized waste treatment facility after the radioactivity in effluent is checked within power block. The liquid effluent from centralized waste treatment facility will be discharged by way of discharge canal in order to be in the sufficient condition. As a result of investigating the radiation monitoring design in accordance with 20 provisions by Korean Regulatory Authority, each effluent radiation monitoring with 20 provisions by Korean Regulatory Authority, each effluent radiation monitoring design satisfies the regulatory guideline. In relation to sampling and analyses, most systems satisfy the regulatory guideline except for some effluents from turbine building. And, though sampling and analyses are performed after radioactivity is monitored at each system in turbine building, these exceptions in turbine building effluents are expected to cause no significant problems because radioactivity is monitored by direct or indirect methods prior to release from turbine building. Integrated monitoring on liquid effluent from the centralized waste water treatment facility is not necessary because radiation monitoring, sampling and analyses on each system within power block are performed, and operational effectiveness compared with cost according to adding the radiation monitoring equipment is too low. So, whether the radiation monitoring in this effluent is reflected on design or not is planned to be determined through discussion with regulatory authority

High performance light water reactor

International Nuclear Information System (INIS)

Squarer, D.; Schulenberg, T.; Struwe, D.; Oka, Y.; Bittermann, D.; Aksan, N.; Maraczy, C.; Kyrki-Rajamaeki, R.; Souyri, A.; Dumaz, P.

2003-01-01

The objective of the high performance light water reactor (HPLWR) project is to assess the merit and economic feasibility of a high efficiency LWR operating at thermodynamically supercritical regime. An efficiency of approximately 44% is expected. To accomplish this objective, a highly qualified team of European research institutes and industrial partners together with the University of Tokyo is assessing the major issues pertaining to a new reactor concept, under the co-sponsorship of the European Commission. The assessment has emphasized the recent advancement achieved in this area by Japan. Additionally, it accounts for advanced European reactor design requirements, recent improvements, practical design aspects, availability of plant components and the availability of high temperature materials. The final objective of this project is to reach a conclusion on the potential of the HPLWR to help sustain the nuclear option, by supplying competitively priced electricity, as well as to continue the nuclear competence in LWR technology. The following is a brief summary of the main project achievements:-A state-of-the-art review of supercritical water-cooled reactors has been performed for the HPLWR project.-Extensive studies have been performed in the last 10 years by the University of Tokyo. Therefore, a 'reference design', developed by the University of Tokyo, was selected in order to assess the available technological tools (i.e. computer codes, analyses, advanced materials, water chemistry, etc.). Design data and results of the analysis were supplied by the University of Tokyo. A benchmark problem, based on the 'reference design' was defined for neutronics calculations and several partners of the HPLWR project carried out independent analyses. The results of these analyses, which in addition help to 'calibrate' the codes, have guided the assessment of the core and the design of an improved HPLWR fuel assembly. Preliminary selection was made for the HPLWR scale

Advanced steam cycles for light water reactors. Final report

International Nuclear Information System (INIS)

Mitchell, R.C.

1975-07-01

An appraisal of the potential of adding superheat to improve the overall LWR plant cycle performance is presented. The study assesses the economic and technical problems associated with the addition of approximately 500 0 F of superheat to raise the steam temperature to 1000 0 F. The practicality of adding either nuclear or fossil superheat to LWR's is reviewed. The General Electric Company Boiling Water Reactor (BWR) model 238-732 (BWR/6) is chosen as the LWR starting point for this evaluation. The steam conditions of BWR/6 are representative of LWR's. The results of the fossil superheat portion of the evaluation are considered directly applicable to all LWR's. In spite of the potential of a nuclear superheater to provide a substantial boost to the LWR cycle efficiency, nuclear superheat offers little promise of development at this time. There are difficult technical problems to resolve in the areas of superheat fuel design and emergency core cooling. The absence of a developed high integrity, high temperature fuel for operation in the steam/water environment is fundamental to this conclusion. Fossil superheat offers the potential opportunity to utilize fossil fuel supplies more efficiently than in any other mode of central station power generation presently available. Fossil superheat topping cycles evaluated included atmospheric fluidized beds (AFB), pressurized fluidized beds, pressurized furnaces, conventional furnaces, and combined gas/steam turbine cycles. The use of an AFB is proposed as the preferred superheat furnace. Fossil superheat provides a cycle efficiency improvement for the LWR of two percentage points, reduces heat rejection by 15 percent per kWe generated, increases plant electrical output by 54 percent, and burns coal with an incremental net efficiency of approximately 40 percent. This compares with a net efficiency of 36--37 percent which might be achieved with an all-fluidized bed fossil superheat plant design

Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

International Nuclear Information System (INIS)

Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y.

2001-01-01

The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 ∼ 10 -V at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

Spent fuel data base: commercial light water reactors

International Nuclear Information System (INIS)

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel

Advanced Solid State Lighting for Human Evaluation

Data.gov (United States)

National Aeronautics and Space Administration — Lighting intensity and color have a significant impact on human circadian rhythms.  Advanced solid state lighting was developed for the Advanced Exploration System...

Thermophysical properties database of materials for light water reactors and heavy water reactors. Final report of a coordinated research project 1999-2005

International Nuclear Information System (INIS)

2006-06-01

The IAEA Coordinated Research Project (CRP) on the Establishment of a Thermo-physical Properties Database for Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs) started in 1999. It was included in the IAEA's Nuclear Power Programme following endorsement in 1997 by the IAEA's Technical Working Groups on Advanced Technologies for LWRs and HWRs (the TWG-LWR and the TWG-HWR). Furthermore, the TWG on Fuel Performance and Technology (TWG-FPT) also expressed its support. This CRP was conducted as a joint task within the IAEA's project on technology development for LWRs and HWRs in its nuclear power programme. Improving the technology for nuclear reactors through better computer codes and more accurate materials property data can contribute to improved economics of future plants by helping to remove the need for large design margins, which are currently used to account for limitations of data and methods. Accurate representations of thermo-physical properties under relevant temperature and neutron fluence conditions are necessary for evaluating reactor performance under normal operation and accident conditions. The objective of this CRP was to collect and systematize a thermo-physical properties database for light and heavy water reactor materials under normal operating, transient and accident conditions and to foster the exchange of non-proprietary information on thermo-physical properties of LWR and HWR materials. An internationally available, peer reviewed database of properties at normal and severe accident conditions has been established on the Internet. This report is intended to serve as a useful source of information on thermo-physical properties data for water cooled reactor analyses. The properties data have been initially stored in the THERSYST data system at the University of Stuttgart, Germany, which was subsequently developed into an internationally available Internet database named THERPRO at Hanyang University, Republic of Korea

Future nuclear power generation

International Nuclear Information System (INIS)

Mosbah, D.S.; Nasreddine, M.

2006-01-01

The book includes an introduction then it speaks about the options to secure sources of energy, nuclear power option, nuclear plants to generate energy including light-water reactors (LWR), heavy-water reactors (HWR), advanced gas-cooled reactors (AGR), fast breeder reactors (FBR), development in the manufacture of reactors, fuel, uranium in the world, current status of nuclear power generation, economics of nuclear power, nuclear power and the environment and nuclear power in the Arab world. A conclusion at the end of the book suggests the increasing demand for energy in the industrialized countries and in a number of countries that enjoy special and economic growth such as China and India pushes the world to search for different energy sources to insure the urgent need for current and anticipated demand in the near and long-term future in light of pessimistic and optimistic outlook for energy in the future. This means that states do a scientific and objective analysis of the currently available data for the springboard to future plans to secure the energy required to support economy and welfare insurance.

Global development of advanced nuclear power plants, and related IAEA activities

International Nuclear Information System (INIS)

2006-09-01

Renewed interest in the potential of nuclear energy to contribute to a sustainable worldwide energy mix is underlining the IAEA's statutory role in fostering the peaceful uses of nuclear energy, in particular the need for effective exchanges of information and collaborative research and technology development among Member States on advanced nuclear power technologies deployable in the near term as well as in the longer term. For applications in the medium to longer term, with rising expectations for the role of nuclear energy in the future, technological innovation has become a strong focus of nuclear power technology developments by many Member States. To meet Member States' needs, the IAEA conducts activities to foster information exchange and collaborative research and development in the area of advanced nuclear reactor technologies. These activities include coordination of collaborative research, organization of international information exchange, and analyses of globally available technical data and results, with a focus on reducing nuclear power plant capital costs and construction periods while further improving performance, safety and proliferation resistance. In other activities, evolutionary and innovative advances are catalyzed for all reactor lines such as advanced water cooled reactors, high temperature gas cooled reactors, liquid metal cooled reactors and accelerator driven systems, including small and medium sized reactors. In addition, there are activities related to other applications of nuclear energy such as seawater desalination, hydrogen production, and other process heat applications. This brochure summarizes the worldwide status and the activities related to advanced nuclear power technology development and related IAEA activities. It includes a list of the collaborative research and development projects conducted by the IAEA, as well as of the status reports and other publications produced

Handbook of advanced nuclear hydrogen safety. 1st edition

International Nuclear Information System (INIS)

Hino, Ryutaro; Takegami, Hiroaki; Ogawa, Toru

2017-03-01

In the aftermath of the Fukushima nuclear accident, safety measures against hydrogen in severe accident has been recognized as a serious technical problem in Japan. Therefore, efforts have begun to form a common knowledge base between nuclear engineers and experts on combustion and explosion, and to secure and improve future nuclear energy safety. As one of such activities, we have prepared the 'Handbook of Advanced Nuclear Hydrogen Safety'. A handbook committee consisting of Japanese experts in the fields of nuclear and combustion-explosion in universities, nuclear companies, electric companies and research institutes was established in 2012. The objective and consents of the handbook were determined, and the outline of the contents was decided. The concepts of the handbook are as follows: to show advanced nuclear hydrogen safety technologies that nuclear engineers should understand, to show hydrogen safety points to make combustion-explosion experts cooperate with nuclear engineers, to expand information on water radiolysis considering the situation from just after the Fukushima accidents and to the waste management necessary for decommissioning after the accident etc. Many experts have participated to manuscript preparation, which was the first step of forming a hydrogen community across the boundaries of fields. The hydrogen community is expected to grow along with its improvement to the knowledge base on nuclear hydrogen safety. (author)

Importance of Advanced Planning of Manufacturing for Nuclear Industry

Directory of Open Access Journals (Sweden)

Shykinov Nick

2016-06-01

Full Text Available In the context of energy demands by growing economies, climate changes, fossil fuel pricing volatility, and improved safety and performance of nuclear power plants, many countries express interest in expanding or acquiring nuclear power capacity. In the light of the increased interest in expanding nuclear power the supply chain for nuclear power projects has received more attention in recent years. The importance of the advanced planning of procurement and manufacturing of components of nuclear facilities is critical for these projects. Many of these components are often referred to as long-lead items. They may be equipment, products and systems that are identified to have a delivery time long enough to affect directly the overall timing of a project. In order to avoid negatively affecting the project schedule, these items may need to be sourced out or manufactured years before the beginning of the project. For nuclear facilities, long-lead items include physical components such as large pressure vessels, instrumentation and controls. They may also mean programs and management systems important to the safety of the facility. Authorized nuclear operator training, site evaluation programs, and procurement are some of the examples. The nuclear power industry must often meet very demanding construction and commissioning timelines, and proper advanced planning of the long-lead items helps manage risks to project completion time. For nuclear components there are regulatory and licensing considerations that need to be considered. A national nuclear regulator must be involved early to ensure the components will meet the national legal regulatory requirements. This paper will discuss timing considerations to address the regulatory compliance of nuclear long-lead items.

Estimation, comparison, and evaluation of advanced fission power reactor generation costs

International Nuclear Information System (INIS)

Waddell, J.D.

1977-01-01

The study compares the high-temperature gas-cooled reactor (HTGR), the gas-cooled fast reactor (GCFR), the molten-salt breeder reactor (MSBR), the light water breeder reactor (LWBR), and the heavy water reactor (HWR) with proposed light water reactors (LWR) and liquid-metal fast breeder reactors (LMFBR). The relative electrical generation costs, including the effects of the introduction of advanced reactor fuel cycles into the U.S. nuclear power economy, were projected through the year 2030. The study utilized the NEEDS computer code which is a simulation of the U.S. nuclear power economy. The future potential electrical generation costs and cumulative consumption of uranium ore were developed using characterizations of the advanced systems. The reactor-fuel cycle characterizations were developed from literature reviews and personal discussions with the proponents of the various systems. The study developed a ranking of the concepts based on generation costs and uranium consumption

Light and heavy water replacing system in reactor container

International Nuclear Information System (INIS)

Miyamoto, Keiji.

1979-01-01

Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

Special issue on the "Consortium for Advanced Simulation of Light Water Reactors Research and Development Progress"

Science.gov (United States)

Turinsky, Paul J.; Martin, William R.

2017-04-01

In this special issue of the Journal of Computational Physics, the research and development completed at the time of manuscript submission by the Consortium for Advanced Simulation of Light Water Reactors (CASL) is presented. CASL is the first of several Energy Innovation Hubs that have been created by the Department of Energy. The Hubs are modeled after the strong scientific management characteristics of the Manhattan Project and AT&T Bell Laboratories, and function as integrated research centers that combine basic and applied research with engineering to accelerate scientific discovery that addresses critical energy issues. Lifetime of a Hub is expected to be five or ten years depending upon performance, with CASL being granted a ten year lifetime.

Impact of different moderator ratios with light and heavy water cooled reactors in equilibrium states

International Nuclear Information System (INIS)

Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

2006-01-01

As an issue of sustainable development in the world, energy sustainability using nuclear energy may be possible using several different ways such as increasing breeding capability of the reactors and optimizing the fuel utilization using spent fuel after reprocessing as well as exploring additional nuclear resources from sea water. In this present study the characteristics of light and heavy water cooled reactors for different moderator ratios in equilibrium states have been investigated. The moderator to fuel ratio (MFR) is varied from 0.1 to 4.0. Four fuel cycle schemes are evaluated in order to investigate the effect of heavy metal (HM) recycling. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of SRAC 2000 code using nuclear data library from the JENDL 3.2. The results show a thermal spectrum peak appears for light water coolant and no thermal peak for heavy water coolant along the MFR (0.1 ≤ MFR ≤ 4.0). The plutonium quality can be reduced effectively by increasing the MFR and number of recycled HM. Considering the effect of increasing number of recycled HM; it is also effective to reduce the uranium utilization and to increase the conversion ratio. trans-Plutonium production such as americium (Am) and curium (Cm) productions are smaller for heavy water coolant than light water coolant. The light water coolant shows the feasibility of breeding when HM is recycled with reducing the MFR. Wider feasible area of breeding has been obtained when light water coolant is replaced by heavy water coolant

Bounding analysis of containment of high pressure melt ejection in advanced light water reactors

International Nuclear Information System (INIS)

Additon, S.L.; Fontana, M.H.; Carter, J.C.

1990-01-01

This paper reports on the loadings on containment due to direct containment heating (DCH) as a result of high pressure melt ejection (HPME) in advanced light water reactors (ALWR) which were estimated using conservative, bounding analyses. The purpose of the analyses was to scope the magnitude of the possible loadings and to indicate the performance needed from potential mitigation methods, such as a cavity configuration that limits energy transfer to the upper containment volume. Analyses were performed for three cases which examined the effect of availability of high pressure reactor coolant system water at the time of reactor vessel melt through and the effect of preflooding of the reactor cavity. The amount of core ejected from the vessel was varied from 100% to 0% for all cases. Results indicate that all amounts of core debris dispersal could be accommodated by the containment for the case where the reactor cavity was preflooded. For the worst case, all the energy from in-vessel hydrogen generation and combustion plus that from 45% of the entire molten core would be required to equilibrate with the containment upper volume in order to reach containment failure pressure

Enhancement of fatigue crack growth rates in pressure boundary materials due to light-water-reactor environments

International Nuclear Information System (INIS)

VanDerSluys, W.A.; Emanuelson, R.H.

1988-01-01

The high level of reliability required of the primary-coolant pressure boundary in a nuclear reactor system leads to a continuing interest in the interaction among the coolant, pressure boundary materials, and service loadings. One area of concern involves the possible enhancement of the growth rate of fatigue cracks due to the coolant. Advances have occurred recently toward a better understanding of the variables influencing the material/environment interactions and methods of addressing this interaction. Sulfur now appears to be one of the principal agents responsible for the observed enhancement of the fatigue crack growth rates in light-water-reactor (LWR) environments. This paper presents the results of investigations on the effect of sulfur in the steel, bulk water environment, and at the crack tip

Database structure and file layout of Nuclear Power Plant Database. Database for design information on Light Water Reactors in Japan

International Nuclear Information System (INIS)

Yamamoto, Nobuo; Izumi, Fumio.

1995-12-01

The Nuclear Power Plant Database (PPD) has been developed at the Japan Atomic Energy Research Institute (JAERI) to provide plant design information on domestic Light Water Reactors (LWRs) to be used for nuclear safety research and so forth. This database can run on the main frame computer in the JAERI Tokai Establishment. The PPD contains the information on the plant design concepts, the numbers, capacities, materials, structures and types of equipment and components, etc, based on the safety analysis reports of the domestic LWRs. This report describes the details of the PPD focusing on the database structure and layout of data files so that the users can utilize it efficiently. (author)

IRIS - Generation IV Advanced Light Water Reactor for Countries with Small and Medium Electricity Grids

International Nuclear Information System (INIS)

Carelli, M. D.

2002-01-01

An international consortium of industry, laboratory, university and utility establishments, led by Westinghouse, is developing a Generation IV Reactor, International Reactor Innovative and Secure (IRIS). IRIS is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., fuel cycle sustainability, enhanced reliability and safety, and improved economics. It features innovative, advanced engineering, but it does not require new technology development since it relies on the proven technology of light water reactors. This paper presents the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and four-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. The path forward for possible future extension to a eight-year cycle will be also discussed. IRIS has a large potential worldwide market because of its proven technology, modularity, low financing, compatibility with existing grids and very limited infrastructure requirements. It is especially appealing to developing countries because of ease of operation and because its medium power is more adaptable to smaller grids. (author)

Size control synthesis and characterization of ZnO nanoparticles and its application as ZnO-water based nanofluid in heat transfer enhancement in light water nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Sharma, Deepak; Pandey, Krishna Murari [National Institute of Technology Silchar, Assam (India). Dept. of Mechanical Engineering

2017-03-15

A novel and facile approach for size-tunable synthesis of ZnO nanoparticle (NPs) is reported. Size-tuning was attained by using PEG (polyethylene glycol) of molecular weights 400 and 4000. ZnO NPs was synthesized using homogeneous chemical precipitation followed by hydrothermal. Here triethylamine (TEA) was used as a hydroxylating agent. As-synthesized ZnO NPs were characterized by powder X-ray diffraction (XRD), transmission electron microscopy (TEM) and Energy Dispersive Spectroscopy (EDS) analysis. Synthesized ZnO nanoparticle was used for the preparation of ZnO-water based nanofluid and its application in heat transfer enhancement in light water nuclear reactor. In this work, ZnO-water based nanofluid of different volume concentration (1%, 2% and 3%) and particle size of 10 nm and 20 nm is used for enhancement in heat transfer in annular channel by using two phase approach. The particle size of 10 nm gives better result for enhancing the heat transfer rate in comparison to 20 nm particle size in nuclear reactor.

77 FR 15812 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

Science.gov (United States)

2012-03-16

... Systems for Light-Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... Feedwater Systems for Light- Water Reactors.'' DG-1265 is proposed revision 2 of Regulatory Guide 1.68.1... Plants,'' dated January 1977. This regulatory guide is being revised to: (1) expand the scope of the...

Technology programs in support of advanced light water reactor plants: Construction

International Nuclear Information System (INIS)

Eichen, E.P.

1989-10-01

Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule duration, and to simplify design. This document discusses construction approaches. 77 refs., 5 figs., 6 tabs

Nuclear imaging: Advances and trends

International Nuclear Information System (INIS)

Herk, G. van

1986-01-01

In this article, nuclear imaging instruments that are likely to be of interest to the nuclear medicine community of developing countries are emphasized. The advances, trends, developments, and future directions in the field of nuclear imaging are mentioned

Guidebook on quality control of mixed oxides and gadolinium bearing fuels for light water reactors

International Nuclear Information System (INIS)

1991-02-01

Under the coverage of an efficient quality assurance system, quality control in nuclear fuel fabrication is an essential element to assure the reliable performance of all its components in service. Incentives to increase fuel performance, by extending reactor cycles or achieving higher burnups and, in some countries to use recycled plutonium in light water reactors (LWRs) necessitated the development of new types of fuels. In the first case, due to higher uranium enrichments, a burnable neutron absorber was integrated to the fuel pellets. Gadolinia was found to form a solid solution with Uranium dioxide and, to present a burnup rate which matches fissile uranium depletion. (U,Gd)O 2 fuels which have been successfully used since the seventies, in boiling water reactors have more recently found an increased utilization, in pressurized water reactors. This amply justifies the publication of this TECDOC to encourage authorities, designers and manufacturers of these types of fuel to establish a more uniform, adapted and effective system of control, thus promoting improved materials reliability and good performance in advanced fuel for light water reactors. The Guidebook is subdivided into four chapters written by different authors. A separate abstract was prepared for each of these chapters. Refs, figs and tabs

Advancing Water Science through Improved Cyberinfrastructure

Science.gov (United States)

Koch, B. J.; Miles, B.; Rai, A.; Ahalt, S.; Band, L. E.; Minsker, B.; Palmer, M.; Williams, M. R.; Idaszak, R.; Whitton, M. C.

2012-12-01

Major scientific advances are needed to help address impacts of climate change and increasing human-mediated environmental modification on the water cycle at global and local scales. However, such advances within the water sciences are limited in part by inadequate information infrastructures. For example, cyberinfrastructure (CI) includes the integrated computer hardware, software, networks, sensors, data, and human capital that enable scientific workflows to be carried out within and among individual research efforts and across varied disciplines. A coordinated transformation of existing CI and development of new CI could accelerate the productivity of water science by enabling greater discovery, access, and interoperability of data and models, and by freeing scientists to do science rather than create and manage technological tools. To elucidate specific ways in which improved CI could advance water science, three challenges confronting the water science community were evaluated: 1) How does ecohydrologic patch structure affect nitrogen transport and fate in watersheds?, 2) How can human-modified environments emulate natural water and nutrient cycling to enhance both human and ecosystem well-being?, 3) How do changes in climate affect water availability to support biodiversity and human needs? We assessed the approaches used by researchers to address components of these challenges, identified barriers imposed by limitations of current CI, and interviewed leaders in various water science subdisciplines to determine the most recent CI tools employed. Our preliminary findings revealed four areas where CI improvements are likely to stimulate scientific advances: 1) sensor networks, 2) data quality assurance/quality control, 3) data and modeling standards, 4) high performance computing. In addition, the full potential of a re-envisioned water science CI cannot be realized without a substantial training component. In light of these findings, we suggest that CI

Technology programs in support of advanced light water reactor plants: Construction

International Nuclear Information System (INIS)

Eichen, E.P.

1987-12-01

Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule durations, and to simplify design. This document discusses successes and problems in construction. 49 refs., 16 figs., 8 tabs

Light water reactor (LWR) innovation needs in the United States: The Massachusetts Institute of Technology LWR innovation project

International Nuclear Information System (INIS)

Golay, M.W.

1988-01-01

A major effort under way within the Massachusetts Institute of Technology (MIT) Engineering School is focused on the contributions that technology innovation can make in revitalizing nuclear power in the United States. A principal component of this effort is a project to improve the designs of the next generation of light water reactors (LWRs) with emphasis on achieving improved capacity factors and safety, and reducing the construction duration. The motivation for this overall effort is to prevent the nuclear option from being unnecessarily lost by being available only in uneconomic configurations. In considering how to advance this effort, the authors focused on refining the designs of new reactors because this is the area where the greatest opportunities for improvements exist

Down-selection of candidate alloys for further testing of advanced replacement materials for LWR core internals

Energy Technology Data Exchange (ETDEWEB)

Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States). Applied Physics Program; Leonard, Keith J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Tan, Lizhen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) Light Water Reactor Sustainability Program to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to identify and develop advanced alloys with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024.

In-operation inspection technology development 'development of a rational maintenance management method for light-water reactor plant'

International Nuclear Information System (INIS)

Matsumoto, K.; Sanoh, J.; Uhara, Y.; Takeshima, K.; Tani, M.; O'Shima, E.

2001-01-01

In 1985, the Japanese national project named 'In-Operation Inspection Technology Development (IOI)' was initiated, as a part of the activities for advancing the LWR(light water reactor)technology in Japan. This project developed the techniques for in-operation monitoring and detecting of early anomalies of nuclear power equipment such as rotating machines, valves and piping. Further, the estimation systems for diagnosing and predicting a degradation rate of these items of equipment were constructed. Based on these results, a new maintenance management technology was constructed. This paper describes the outline of the new maintenance management concept. (authors)

Proceedings of 2017 international congress on advances in nuclear power plants (ICAPP2017)

International Nuclear Information System (INIS)

2017-04-01

The International Congress on Advances in Nuclear Power Plants (ICAPP) provides a forum for leaders of the nuclear industry to exchange information, present results from their work, review the state of the industry, and discuss future directions and needs for the deployment of new nuclear power plant systems around the world. ICAPP will gather industry leaders in several invited lectures in plenary sessions. The theme for ICAPP2017 is 'A New Paradigm in Nuclear Power Safety'. Since the Fukushima Daiichi Accident in 2011, various efforts in improving nuclear safety have been initiated not only in Japan but also in other countries. Decontamination of affected soil and steps toward decommissioning Fukushima Daiichi are proceeding steadily, but many issues to be resolved still remain. Further advances in reactor decommissioning technologies are expected in light of the rising number of old nuclear power plants being closed. The congress also provides an excellent opportunity to discuss these topics. This issue is the collection of 345 papers presented at the entitled meeting. All the 345 papers are indexed individually. (J.P.N.)

Advanced energy system with nuclear reactors as an energy source

International Nuclear Information System (INIS)

Kato, Y.; Ishizuka, T.; Nikitin, K.

2007-01-01

recovery system is also applicable to a fast reactor (FR) with a supercritical CO 2 gas turbine that achieves higher cycle efficiency than conventional sodium cooled FRs with steam turbines. The FR will eliminate problems of conventional FRs related to safety, plant maintenance, and construction costs. The FR consumes efficiently trans-uranium elements (TRU) produced in light water reactors as fuel and reduce long-lived radioactive wastes or environmental loads of long term geological disposal. An Advanced Energy System (AES) with nuclear reactors as an energy source has been proposed which supply electricity and heat to cities. The AES has three objectives: 1. Save energy resources and reduce green house gas emissions, attaining total energy utilization efficiency higher than 85% through waste heat recovery and utilization. 2. Foster a recycling society that produces methane and methanol for fuel cells from waste products of cities and farms. 3. Consume TRU produced in LWRs as fuel for FRs, and reduce long-lived radioactive wastes or environmental loads of long term geological disposal. References 1. Y. Kato, T. Nitawaki and K. Fujima, 'Zero Waste Heat Release Nuclear Cogeneration System, 'Proc. 2003 Intl. Congress on Advanced Nuclear Power Plants (ICAPP'03), Cordoba, Spain, May 4-7, 2003, Paper 3313. 2. Y. Kato, T. Nitawaki and Y. Muto, 'Medium Temperature Carbon Dioxide Gas Turbine Reactor, 'Nucl. Eng. Design, 230, pp. 195-207 (2004). 3. H. N. Tran and Y. Kato, 'New 2 37Np Burning Strategy in a Supercritical CO 2 Cooled Fast Reactor Core Attaining Zero Burnup Reactivity Loss,' Proc. American Nuclear Society's Topical Meeting on Reactor Physics (PHYSOR 2006), Vancouver, British Columbia, Canada, September 10-14, 2006

On Cherenkov light production by irradiated nuclear fuel rods

International Nuclear Information System (INIS)

Branger, E.; Grape, S.; Svärd, S. Jacobsson; Jansson, P.; Sundén, E. Andersson

2017-01-01

Safeguards verification of irradiated nuclear fuel assemblies in wet storage is frequently done by measuring the Cherenkov light in the surrounding water produced due to radioactive decays of fission products in the fuel. This paper accounts for the physical processes behind the Cherenkov light production caused by a single fuel rod in wet storage, and simulations are presented that investigate to what extent various properties of the rod affect the Cherenkov light production. The results show that the fuel properties have a noticeable effect on the Cherenkov light production, and thus that the prediction models for Cherenkov light production which are used in the safeguards verifications could potentially be improved by considering these properties. It is concluded that the dominating source of the Cherenkov light is gamma-ray interactions with electrons in the surrounding water. Electrons created from beta decay may also exit the fuel and produce Cherenkov light, and e.g. Y-90 was identified as a possible contributor to significant levels of the measurable Cherenkov light in long-cooled fuel. The results also show that the cylindrical, elongated fuel rod geometry results in a non-isotropic Cherenkov light production, and the light component parallel to the rod's axis exhibits a dependence on gamma-ray energy that differs from the total intensity, which is of importance since the typical safeguards measurement situation observes the vertical light component. It is also concluded that the radial distributions of the radiation sources in a fuel rod will affect the Cherenkov light production.

Robot vision for nuclear advanced robot

International Nuclear Information System (INIS)

Nakayama, Ryoichi; Okano, Hideharu; Kuno, Yoshinori; Miyazawa, Tatsuo; Shimada, Hideo; Okada, Satoshi; Kawamura, Astuo

1991-01-01

This paper describes Robot Vision and Operation System for Nuclear Advanced Robot. This Robot Vision consists of robot position detection, obstacle detection and object recognition. With these vision techniques, a mobile robot can make a path and move autonomously along the planned path. The authors implemented the above robot vision system on the 'Advanced Robot for Nuclear Power Plant' and tested in an environment mocked up as nuclear power plant facilities. Since the operation system for this robot consists of operator's console and a large stereo monitor, this system can be easily operated by one person. Experimental tests were made using the Advanced Robot (nuclear robot). Results indicate that the proposed operation system is very useful, and can be operate by only person. (author)

Design and screening of nanoprecipitates-strengthened advanced ferritic alloys

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); Chen, Tianyi [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sridharan, K. [Univ. of Wisconsin, Madison, WI (United States); He, Li [Univ. of Wisconsin, Madison, WI (United States)

2016-12-30

Advanced nuclear reactors as well as the life extension of light water reactors require advanced alloys capable of satisfactory operation up to neutron damage levels approaching 200 displacements per atom (dpa). Extensive studies, including fundamental theories, have demonstrated the superior resistance to radiation-induced swelling in ferritic steels, primarily inherited from their body-centered cubic (bcc) structure. This study aims at developing nanoprecipitates strengthened advanced ferritic alloys for advanced nuclear reactor applications. To be more specific, this study aims at enhancing the amorphization ability of some precipitates, such as Laves phase and other types of intermetallic phases, through smart alloying strategy, and thereby promote the crystalline®amorphous transformation of these precipitates under irradiation.

Development status and application prospect of supercritical-pressure light water cooled reactor

International Nuclear Information System (INIS)

Li Manchang; Wang Mingli

2006-01-01

The Supercritical-pressure Light Water Cooled Reactor (SCWR) is selected by the Generation IV International Forum (GIF) as one of the six Generation IV nuclear systems that will be developed in the future, and it is an innovative design based on the existing technologies used in LWR and supercritical coal-fired plants. Technically, SCWR may be based on the design, construction and operation experiences in existing PWR and supercritical coal-fired plants, which means that there is no insolvable technology difficulties. Since PWR technology will be adopted in the near term and medium term projects in China, and considering the sustainable development of the technology, it is an inevitable choice to research and develop the nuclear system of supercritical light water cooled reactor. (authors)

Safety considerations concerning light water reactors in Sweden

International Nuclear Information System (INIS)

Nilsson, T.

1977-01-01

In 1975 the Swedish Nuclear Power Inspectorate was commissioned by the Government to perform a Reactor Safety Study concerning commercial light water reactors. The study will contain an account of: - rules and regulations for reactor designs; - operation experience of the Swedish nuclear power plants with international comparisons; - the development of reactor designs during the last 10 years; - demands and conditions for inspection and inspection methods; - nuclear power plant operation organization; - training of operators; and - the results of research into nuclear safety. The study is scheduled for completion by July 1st, 1977, however, this paper gives a summary of the results of the Reactor Safety Study already available. The paper contains detailed statistics concerning safety related occurrences and reactor scrams in Sweden from July 1st, 1974 until the beginning of 1977

Development of advanced secondary chemistry monitoring system for Korea nuclear power plants

International Nuclear Information System (INIS)

Lee, Sang Hak; Kim, Chung Tae

1997-01-01

Water chemistry control is one of the most important tasks in order to maintain the reliability of plant equipments and extend the operating life of the plant. KEPCO and KOPEC developed a computerized tool for this purpose -ASCMS (advanced secondary chemistry monitoring system) which is able to monitor and diagnose the secondary water chemistry. A prototype system had been installed at KORI 3 nuclear power plant since April 1993 in order to evaluate the system performance. After the implementation of enhancements identified during the testing of the prototype, we have developed the advanced secondary monitoring system, ASCMS which is installed at 5 nuclear power plants and has been under operations since April 1997. The ASCMS comprises PC subsystem designed for data acquisition, data analysis, and data diagnosis. The ASCMS will provide overall information related to steam generator secondary side water chemistry problems and improve plant availability, reduce radiation exposure to workers, and reduce operating and maintenance costs. 6 figs

A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

Energy Technology Data Exchange (ETDEWEB)

Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Liao, Huafei [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-12-01

United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constituted a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.

Current status of light water reactor and Hitachi's technical improvements for BWR

International Nuclear Information System (INIS)

Miki, Minoru; Ohki, Arahiko.

1984-01-01

Gradual technical improvements in Japan over the years has improved the reliability of light water reactors, and has achieved the highest capacity factor level in the world. Commercial operation of Fukushima 2-2 (1,100 MW) of the Tokyo Electric Power Co. was started in February, 1984, as the first standardized BWR base plant, ushering in a new age of domestic light water reactor technology. The ABWR (1,300 MW class) has been developed as Japan's next generation light water reactor, with construction aimed at the latter half of the 1980's. Hitachi's extensive efforts range from key nuclear equipment to various related robots, directed at improving safety, reliability, and the capacity factor, while reducing radiation exposure. This paper presents an outline of Hitachi's participation in the light water reactor's improvement and standardization, and the current status of our role in the international cooperation plan for the ABWR. (author)

Supercritical Water Nuclear Steam Supply System: Innovations In Materials, Neutronics and Thermal-Hydraulics

International Nuclear Information System (INIS)

Anderson, Mark; Corradini, M.L.; Sridharan, K.; Wilson, P.; Cho, D.; Kim, T.K.; Lomperski, S.

2004-01-01

In the 1990's supercritical light-water reactors were considered in conceptual designs. A nuclear reactor cooled by supercritical waster would have a much higher thermal efficiency with a once-through direct power cycle, and could be based on standardized water reactor components (light water or heavy water). The theoretical efficiency could be improved by more than 33% over that of other water reactors and could be simplified with higher reliability; e.g., a boiling water reactor without steam separators or dryers

Assessment of EPRI water chemistry guidelines for new nuclear power plants

International Nuclear Information System (INIS)

Reid Richard; Kim Karen; McCree, Anisa; Eaker, Richard; Sawochka, Steve; Giannelli, Joe

2012-09-01

Water chemistry control technologies for nuclear power plants have been significantly enhanced over the past few decades to improve material and equipment reliability and fuel performance, and to minimize radionuclide production and transport. Chemistry Guidelines have been developed by the Electric Power Research Institute (EPRI) for currently operating plants and have been intermittently revised over the past twenty-five years for the protection of systems and components and for radiation management. As new plants are being designed for improved safety and increased power production, it is important to ensure that the designs consider implementation of state-of-the-art, industry developed water chemistry controls. In parallel, the industry will need to consider and update water chemistry guidelines as well as plant startup and operational strategies based on the advanced plant designs. EPRI has performed assessments of water chemistry control guidance or assumptions provided in design and licensing documents for several advanced plant designs. These designs include: Westinghouse AP1000 Pressurized Water Reactor AREVA US-EPR Pressurized Water Reactor Mitsubishi Nuclear Energy Systems/Mitsubishi Heavy Industries Advanced Pressurized Water Reactor Korea Hydro and Nuclear Power APR1400 Pressurized Water Reactor Toshiba Advanced Boiling Water Reactor (ABWR) General Electric-Hitachi Economic Simplified Boiling Water Reactor (ESBWR) The intent of these assessments was to identify key design differences in each of the new plant designs relative to the current operating fleet and to identify differences in water chemistry specifications or design assumptions provided in design and licensing documents for the plants in comparison to current EPRI Water Chemistry Guidelines. This paper provides a summary of the key results of these assessments. The fundamental design and operation of the advanced plants is similar to the currently operating fleet. As such, the new plants are

Technology programs in support of advanced light water reactor plants: Construction

International Nuclear Information System (INIS)

Eichen, E.P.

1989-01-01

Under Contract No. AC03-86SF16565, Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule durations, and to simplify design. This document discusses design requirements. 36 refs., 57 figs., 56 tabs

Advanced chemistry management system for nuclear power plants

International Nuclear Information System (INIS)

Maeda, Katsuji; Kobayashi, Yasuhiro; Nagasawa, Katsumi

2000-01-01

Chemistry control in a boiling water reactor (BWR) plant has a close relationship with radiation field buildup, fuel reliability, integrity of plant components and materials, performance of the water treatment systems and radioactive waste generation. Chemistry management in BWR plants has become more important in order to maintain and enhance plant reliability. Adequate chemistry control and management are also essential to establish, maintain, and enhance plant availability. For these reasons, we have developed the advanced chemistry management system for nuclear power plants in order to effectively collect and evaluate a large number of plant operating and chemistry data. (author)

NEPTUNE: a modular system for light-water reactor calculation

International Nuclear Information System (INIS)

Bouchard, J.; Kanevoky, A.; Reuss, P.

1975-01-01

A complete modular system of light water reactor calculations has been designed. It includes basic nuclear data processing, the APOLLO phase: transport calculations for cells, multicells, fuel assemblies or reactors, the NEPTUNE phase: reactor calculations. A fuel management module, devoted to the automatic determination of the best shuffling strategy is included in NEPTUNE [fr

Light water ultra-safe plant concept

International Nuclear Information System (INIS)

Klevans, E.

1989-01-01

Since the accident at Three Mile Island (TMI), Penn State Nuclear Engineering Department Faculty and Staff have considered various methods to improve already safe reactor designs and public perception of the safety of Nuclear Power. During 1987 and 1988, the Department of Energy provided funds to the Nuclear Engineering Department at Penn State to investigate a plant reconfiguration originated by M.A. Schultz called ''The Light Water Ultra-Safe Plant Concept''. This report presents a final summary of the project with references to several masters' theses and addendum reports for further detail. The two year research effort included design verification with detailed computer simulation of: (a) normal operation characteristics of the unique pressurizing concept, (b) severe transients without loss of coolant, (c) combined primary and secondary system modeling, and (d) small break and large break loss of coolant accidents. Other studies included safety analysis, low power density core design, and control system design to greatly simplify the control room and required operator responses to plant upset conditions. The overall conclusion is that a reconfigured pressurized water reactor can achieve real and perceived safety improvements. Additionally, control system research to produce greatly simplified control rooms and operator requirements should be continued in future projects

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

International Nuclear Information System (INIS)

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

2012-01-01

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

Study of reverse osmosis applicability to light water reactor radwaste processing

International Nuclear Information System (INIS)

Markind, J.; Van Tran, T.

1978-12-01

Objectives were to collect and evaluate documented performance data of existing reverse-osmosis/ultrafiltration processes utilized for treating low-level liquid radioactive wastes, originating from light-water-reactor (LWR) nuclear power plants. Relevant information was collected by communication both written and verbal with membrane experts known to be active in the nuclear industry, and by conducting manual and computer searches. The generated information was evaluated on the basis of membrane performance characteristics relevant to nuclear engineering system analysis. 39 figures, 34 tables

Critical heat flux and flow instability in an advanced light water reactor

International Nuclear Information System (INIS)

Dae-Hyun Hwang; Kyong-Won Seo; Chung-Chan Lee; Sung-Kyun Zee

2005-01-01

Full text of publication follows: An advanced light water reactor concept has been continuously studied in KAERI with an output in the range of about 60 to 300 MW th . The reactor is purposed to be utilized as an energy source for seawater desalination as well as small scale power generation. In order to achieve the intrinsic safety and enhanced operational flexibility, some specific design considerations such as low power density and soluble boron free operation have been incorporated in the multiple-parallel-channel type reactor core. The low power density can be achieved by adopting fuel assemblies with tightly spaced non-square lattice rod array. The allowable core operating region should be primarily limited by the two design parameters; the critical heat flux(CHF) and the flow instabilities in the multiple parallel fuel assembly channels. The characteristics of CHF and flow instability have been investigated through experimental and analytical works. The CHF prediction model was established on the basis of experimental data obtained from 19-rod test bundles. The CHF experiments have been conducted for various test bundles with different heated lengths, uniform and non-uniform radial and axial power distributions, water and Freon as the working fluids, and different number of unheated rods. The parametric ranges of CHF experiments covers the pressure from 6 to 18 MPa, the mass flux from 150 to 2000 kg/m 2 /s, and the inlet subcooling from 10 to 120 deg. C. The flow instabilities due to density wave oscillations were investigated by conducting experiments with two parallel channels under the pressure ranges from 6 to 16 MPa. The parametric behavior of flow instability was examined for the test sections with different lengths of adiabatic risers, different axial power shapes, different inlet restrictions, and different channel cross sections. The stability boundary was experimentally determined by increasing channel inlet temperature or reducing the flow rate

Nuclear reactor design

CERN Document Server

2014-01-01

This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

Non-proliferation and advances in nuclear science

International Nuclear Information System (INIS)

Iyengar, P.K.

1995-01-01

So far, the non-proliferation treaty (NPT) has concentrated on safeguard regimes based on technologies relating to the production of uranium and plutonium in nuclear reactors, and on their potential diversion for use in nuclear weapons. As nuclear science advances, however, nuclear technology both peaceful and for weapons will change, and for the NPT to remain relevant, it must reflect these changes. At this juncture, when the NPT is coming up for review in a year's time, it is important for physicists to take a fresh look at recent advances in nuclear science, and inform the policy-makers and the public at large about their potential for impacting nuclear technology in the future. In this article a few such advances are highlighted and their implications for the NPT are considered. (author). 4 refs

Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

International Nuclear Information System (INIS)

2012-03-01

The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

Advances in water chemistry control for BWRs and PWRs

International Nuclear Information System (INIS)

Wood, C.J.

1997-01-01

This paper is an overview of the effects of water chemistry developments on the current operation of nuclear power plants in the United States, and the mitigation of corrosion-related degradation processes and radiation field build-up processes through the use of advanced water chemistry. Recent modifications in water chemistry to control and reduce radiation fields are outlined, including revisions to the EPRI water chemistry guidelines for BWRs and PWR primary and secondary systems. The change from a single water chemistry specification for all plants to a set of options, from which a plant-specific chemistry programme can be defined, is described. (author)

A retrospective analysis of funding and focus in US advanced fission innovation

Science.gov (United States)

Abdulla, A.; Ford, M. J.; Morgan, M. G.; Victor, D. G.

2017-08-01

Deep decarbonization of the global energy system will require large investments in energy innovation and the deployment of new technologies. While many studies have focused on the expenditure that will be needed, here we focus on how government has spent public sector resources on innovation for a key carbon-free technology: advanced nuclear. We focus on nuclear power because it has been contributing almost 20% of total US electric generation, and because the US program in this area has historically been the world’s leading effort. Using extensive data acquired through the Freedom of Information Act, we reconstruct the budget history of the Department of Energy’s program to develop advanced, non-light water nuclear reactors. Our analysis shows that—despite spending 2 billion since the late 1990s—no advanced design is ready for deployment. Even if the program had been well designed, it still would have been insufficient to demonstrate even one non-light water technology. It has violated much of the wisdom about the effective execution of innovative programs: annual funding varies fourfold, priorities are ephemeral, incumbent technologies and fuels are prized over innovation, and infrastructure spending consumes half the budget. Absent substantial changes, the possibility of US-designed advanced reactors playing a role in decarbonization by mid-century is low.

Light Water Reactor Sustainability Program: Digital Technology Business Case Methodology Guide

Energy Technology Data Exchange (ETDEWEB)

Thomas, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lawrie, Sean [ScottMadden, Inc., Raleigh, NC (United States); Hart, Adam [ScottMadden, Inc., Raleigh, NC (United States); Vlahoplus, Chris [ScottMadden, Inc., Raleigh, NC (United States)

2014-09-01

The Department of Energy’s (DOE’s) Light Water Reactor Sustainability Program aims to develop and deploy technologies that will make the existing U.S. nuclear fleet more efficient and competitive. The program has developed a standard methodology for determining the impact of new technologies in order to assist nuclear power plant (NPP) operators in building sound business cases. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway is part of the DOE’s Light Water Reactor Sustainability (LWRS) Program. It conducts targeted research and development (R&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: (1) to ensure that legacy analog II&C systems are not life-limiting issues for the LWR fleet and (2) to implement digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Resolving long-term operational concerns with the II&C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation’s energy and environmental security. The II&C Pathway is conducting a series of pilot projects that enable the development and deployment of new II&C technologies in existing nuclear plants. Through the LWRS program, individual utilities and plants are able to participate in these projects or otherwise leverage the results of projects conducted at demonstration plants. Performance advantages of the new pilot project technologies are widely acknowledged, but it has proven difficult for utilities to derive business cases for justifying investment in these new capabilities. Lack of a business case is often cited by utilities as a barrier to pursuing wide-scale application of digital technologies to nuclear plant work activities. The decision to move forward with funding usually hinges on

Study of plutonium disposition using existing GE advanced Boiling Water Reactors

Energy Technology Data Exchange (ETDEWEB)

1994-06-01

The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

Study of plutonium disposition using existing GE advanced Boiling Water Reactors

International Nuclear Information System (INIS)

1994-01-01

The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ''Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs

A parametric thermohydraulic study an advanced pressurized light water reactor with a tight fuel rod lattice

International Nuclear Information System (INIS)

Dalle Donne, M.; Hame, W.

1982-12-01

A parametric thermohydraulic study for an Advanced Pressurized Light Water Reactor (APWR) with a tight fuel rod lattice has been performed. The APWR improves the uranium utilisation. The APWR core should be placed in a modern German PWR plant. Within this study about 200 different reactors have been calculated. The tightening of the fuel rod lattice implies a decrease of the net electrical output of the plant, which is greater for the heterogeneous reactor than for the homogeneous reactor. APWR cores mean higher core pressure drops and higher water velocities in the core region. The cores tend to be shorter and the number of fuel rods to be higher than for the PWR. At the higher fuel rod pitch to diameter ratios (p/d) the DNB limitation is more stringent than the limitation on the fuel rod linear rating given by the necessity of reflooding after a reactor accident. The contrary is true for the lower p/d ratios. Subcooled boiling in the highest rated coolant channels occurs for the most of the calculated reactors. (orig.) [de

Light Water Reactor Generic Safety Issues Database (LWRGSIDB). User's manual

International Nuclear Information System (INIS)

1999-01-01

resolved in other plants and which can be used in reassessing the safety of individual operating plants. The IAEA-TECDOC-1044, Generic Safety Issues for Nuclear Power Plants with Light Water Reactors and Measures Taken for Their Resolution (September 1998), is a compilation of such safety issues which is based on broad international experience. This compilation is one element within the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. It is a compilation not only of the generic safety issues identified in nuclear power plants but also, in almost all cases, the measures taken to resolve these issues. The safety issues, which are generic in nature with regard to light water reactors (LWRs), and the measures taken for their resolution, are intended for use as a reference in the reassessment of the safety of operating plants.The information contained in the main body of the TECDOC has been used to establish a database. This database has search, query and report functions. This information is thus available in an electronic form which can be selectively queried and with which reports can be produced according to the requirements of the user. The database also enables the IAEA to update the data periodically on the basis of information made available by Member States

Controllability studies for an advanced CANDU boiling light water reactor

International Nuclear Information System (INIS)

Lepp, R.M.; Hinds, H.W.

1976-12-01

Bulk controllability studies carried out as part of a conceptual design study of a 1200 MWe CANDU boiling-light-water reactor fuelled with U 235 - or Pu-enriched uranium oxide are outlined. The concept, the various models developed for its simulation on a hybrid computer and the perturbations used to test system controllability, are described. The results show that this concept will have better bulk controllability than similar CANDU-BLW reactors fuelled with natural uranium. (author)

EPR (European Pressurized water Reactor) The advanced nuclear reactor

International Nuclear Information System (INIS)

2004-01-01

Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21. century, which puts the emphasis on sustainable development. The EPR is the only 3. generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR was developed by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, a subsidiary of AREVA and Siemens. EDF and the major German electricity companies played an active part in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. The project had three objectives: meet the requirements of European utilities, comply with the safety standards laid down by the French safety authority for future pressurized water reactors, in concert with its German counterpart, and make nuclear energy even more competitive than energy generated using fossil fuels. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. This document presents the main characteristics of the EPR, and in particular the additional measures to prevent the occurrence of events likely to damage the core, the leak-tight containment, the measures to reduce the exposure of operating and maintenance personnel, the solutions for an even greater protection of the environment. The foreseen development of the EPR in France and abroad (Finland, China, the United States) is summarized

Special issue on ageing management and long term operation of light water reactors

International Nuclear Information System (INIS)

Kashima, Koichi; Kanno, Masanori; Kimura, Itsurou

2008-01-01

Ageing management and long term operation of light water reactors became important in Japan and relevant research on physical and materials ageing has been carried out among organizations of government, academia and industry and establishment of technical standards and guideline based on the results is under way. The Japan Energy Policy Institute (JEPI) issued a special number discussing this theme, which consisted of ten reports of experts describing these activities. Main topics were ageing evaluation of reactor components due to neutron irradiation embrittlement and stress corrosion cracking, regulatory evaluation of deteriorations due to ageing with technical information basis, technology development in the area of inspection/monitoring, ageing evaluation and preventive maintenance/repairs, nuclear power plant life management of electric utilities, and advancement of reactor maintenance and inspection. (T. Tanaka)

Light water breeder reactor using a uranium-plutonium cycle

International Nuclear Information System (INIS)

Radkowsky, A.; Chen, R.

1990-01-01

This patent describes a light water receptor (LWR) for breeding fissile material using a uranium-plutonium cycle. It comprises: a prebreeder section having plutonium fuel containing a Pu-241 component, the prebreeder section being operable to produce enriched plutonium having an increased Pu-241 component; and a breeder section for receiving the enriched plutonium from the prebreeder section, the breeder section being operable for breeding fissile material from the enriched plutonium fuel. This patent describes a method of operating a light water nuclear reactor (LWR) for breeding fissile material using a uranium-plutonium cycle. It comprises: operating the prebreeder to produce enriched plutonium fuel having an increased Pu-241 component; fueling a breeder section with the enriched plutonium fuel to breed the fissile material

Department of Energy Nuclear Energy Standards Program

International Nuclear Information System (INIS)

Silver, E.G.

1980-01-01

The policy with respect to the development and use of standards in the Department of Energy (DOE) programs concerned with maintaining and developing the nuclear option for the civilian sector (both in the form of the currently used light water reactors and for advanced concepts including the Liquid Metal Fast Breeder Reactor), is embodied in a Nuclear Standards Policy, issued in 1978, whose perspectives and philosophy are discussed

Status of advanced small pressurized water reactors

International Nuclear Information System (INIS)

Chen Peipei; Zhou Yun

2012-01-01

In order to expand the nuclear power in energy and desalination, increase competitiveness in global nuclear power market, many developed countries with strong nuclear energy technology have realized the importance of Small Modular Reactor (SMR) and initiated heavy R and D programs in SMR. The Advanced Small Pressurized Water Reactor (ASPWR) is characterized by great advantages in safety and economy and can be used in remote power grid and replace mid/small size fossil plant economically. This paper reviews the history and current status of SMR and ASPWR, and also discusses the design concept, safety features and other advantages of ASPWR. The purpose of this paper is to provide an overall review of ASPWR technology in western countries, and to promote the R and D in ASPWR in China. (authors)

Two-phase water hammer in nuclear power plants

International Nuclear Information System (INIS)

Kim, J.H.

1987-01-01

Water hammer events keep recurring in nuclear power plants. In the mid-1970s, water hammer was designated to be an unresolved safety issue (USI A-1) due to its high frequency of occurrence and the severity of the attendant damages. Between 1969 and 1981, a significant number of water hammer incidents (more than 12 events per year) involving BWRs and PWRs have been reported and evaluated. After intensive evaluations of the events, in late 1983, the U.S. Nuclear Regulatory Commission staff concluded that water hammer was not a serious contributor to the degradation of plant safety and it has been taken off the list of the unresolved safety issues. The frequency decreased to around 11 events per year between 1981 and 1985. Nevertheless, 11 events per year are still unacceptable high, especially in light of the financial losses caused by most events. Some events are not required to be reported. The number of the unreported events are estimated to be five to ten times as many as the reported events. The implication is that water hammer in nuclear power plants still needs attention and is a problem that has not been fundamentally resolved

Spent fuel data base: commercial light water reactors. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

Light water reactor safety

CERN Document Server

Pershagen, B

2013-01-01

This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

Status of advanced nuclear reactor development in Korea

International Nuclear Information System (INIS)

Kim, H.R.; Kim, K.K.; Kim, Y.W.; Joo, H.K.

2014-01-01

The Korean nuclear industry is facing new challenges to solve the spent fuel storage problem and meet the needs to diversify the application areas of nuclear energy. In order to provide solutions to these challenges, the Korea Atomic Energy Research Institute (KAERI) has been developing advanced nuclear reactors including a Sodium-cooled Fast Reactor, Very High Temperature Gas cooled Reactor (VHTR), and System-integrated Modular Advanced Reactor (SMART) with substantially improved safety, economics, and environment-friendly features. A fast reactor system is one of the most promising options for a reduction of radioactive wastes. The long-term plan for Advanced SFR development in conjunction with the pyro-process was authorized by the Korean Atomic Energy Commission in 2008. The development milestone includes specific design approval of a prototype SFR by 2020, and the construction of a prototype SFR by 2028. KAERI has been carrying out the preliminary design of a 150MWe SFR prototype plant system since 2012. The development of advanced SFR technologies and the basic key technologies necessary for the prototype SFR are also being carried out. By virtue of high-temperature heat, a VHTR has diverse applications including hydrogen production. KAERI launched a nuclear hydrogen project using a VHTR in 2006, which focused on four basic technologies: the development of design tools, very high-temperature experimental technology, TRISO fuel fabrication, and Sulfur-iodine thermo-chemical hydrogen production technology. The technology development project will be continued until 2017. A conceptual reactor design study was started in 2012 as collaboration between industry and government to enhance the early-launching of the nuclear hydrogen development and demonstration (NHDD) project. The goal of the NHDD project is to design and build a nuclear hydrogen demonstration system by 2030. KAERI has developed SMART which is a small-sized advanced integral reactor with a rated

Calculation of mass flow and steam quality distribution on fuel elements of light-water cooled boiling water nuclear reactors

International Nuclear Information System (INIS)

Hermanns, H.J.

1977-04-01

By the example of light-water cooled nuclear reactors, the state of the calculation methods at disposal for calculating mass flow and steam quality distribution (sub-channel analysis) is indicated. Particular regard was paid to the transport phenomena occurring in reactor fuel elements in the range of two phase flow. Experimentally determined values were compared with recalculations of these experiments with the sub-channel code COBRA; from the results of these comparing calculations, conclusions could be drawn on the suitability of this code for defined applications. Limits of reliability could be determined to some extent. Based on the experience gained and the study of individual physical model concepts, recognized as being important, a sub-channel model was drawn up and the corresponding numerical computer code (SIEWAS) worked out. Experiments made at GE could be reproduced with the code SIEWAS with sufficient accuracy. (orig.) [de

One-phase and two-phase homologous curves for coolant pumps of the pressurized light water nuclear reactors

International Nuclear Information System (INIS)

Santos, G.A. dos.

1990-01-01

The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The single-phase pump characteristics are an essential feature for operational transients studies, for example, the shut-down and start-up of pump. These parameters, in terms of the homologous curves, set up the complete performance of the pump and are input for transients and accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the single-phase and two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

Light-front nuclear shell-model

International Nuclear Information System (INIS)

Johnson, M.B.

1990-01-01

I examine the effects of nuclear structure on high-energy, high-momentum transfer processes, specifically the EMC effect. For pedagogical reasons, a fictitious but simple two-body system consisting of two equal-mass particles interacting in a harmonic oscillator potential has been chosen. For this toy nucleus, I utilize a widely-used link between instant-form and light-front dynamics, formulating nuclear structure and deep-inelastic scattering consistently in the laboratory system. Binding effects are compared within conventional instant-form and light-front dynamical frameworks, with appreciable differences being found in the two cases. 20 refs

Efficiency and cost advantages of an advanced-technology nuclear electrolytic hydrogen-energy production facility

Science.gov (United States)

Donakowski, T. D.; Escher, W. J. D.; Gregory, D. P.

1977-01-01

The concept of an advanced-technology (viz., 1985 technology) nuclear-electrolytic water electrolysis facility was assessed for hydrogen production cost and efficiency expectations. The facility integrates (1) a high-temperature gas-cooled nuclear reactor (HTGR) operating a binary work cycle, (2) direct-current (d-c) electricity generation via acyclic generators, and (3) high-current-density, high-pressure electrolyzers using a solid polymer electrolyte (SPE). All subsystems are close-coupled and optimally interfaced for hydrogen production alone (i.e., without separate production of electrical power). Pipeline-pressure hydrogen and oxygen are produced at 6900 kPa (1000 psi). We found that this advanced facility would produce hydrogen at costs that were approximately half those associated with contemporary-technology nuclear electrolysis: $5.36 versus $10.86/million Btu, respectively. The nuclear-heat-to-hydrogen-energy conversion efficiency for the advanced system was estimated as 43%, versus 25% for the contemporary system.

Advanced Light Source

International Nuclear Information System (INIS)

Sah, R.C.

1983-03-01

The Advanced Light Source (ALS) is a new synchrotron radiation source which has been proposed by Lawrence Berkeley Laboratory. The ALS will be a key component in a major new research facility, the National Center for Advanced Materials. The ALS will consist of an electron linear accelerator, a booster synchrotron, a 1.3-GeV electron storage ring, and a number of photon beam lines. Most or all photon beam lines will originate from wiggler and undulator magnets placed in the 12 long straight sections of the ALS. A very low electron beam emittance will provide photon beams of unsurpassed spectral brilliance from specially-designed undulators, and a high radiofrequency will produce very short pulse lengths

Future of Nuclear Power: NRC emergency preparedness licensing activities agenda

International Nuclear Information System (INIS)

Essig, T.H.

1995-01-01

This talk summary addresses the issue of how future policies of the NRC will affect nuclear power in areas such as construction, emergency preparedness, and licensing. Specific topics covered include the following: Emergent EP licensing issues for operating nuclear Power Plants; 10CFR Part 52 and the process for licensing of Advanced Light Water Reactors (ALWRs); and potential revisions to emergency preparedness programs for future nuclear power plants

Advanced nuclear systems. Review study

International Nuclear Information System (INIS)

Liebert, Wolfgang; Glaser, Alexander; Pistner, Christoph; Baehr, Roland; Hahn, Lothar

1999-04-01

The task of this review study is to from provide an overview of the developments in the field of the various advanced nuclear systems, and to create the basis for more comprehensive studies of technology assessment. In an overview the concepts for advanced nuclear systems pursued worldwide are subdivided into eight subgroups. A coarse examination raster (set pattern) is developed to enable a detailed examination of the selected systems. In addition to a focus on enhanced safety features, further aspects are also taken into consideration, like the lowering of the proliferation risk, the enhancement of the economic competitiveness of the facilities and new usage possibilities (for instance concerning the relaxation of the waste disposal problem or the usage of alternative fuels to uranium). The question about the expected time span for realization and the discussion about the obstacles on the way to a commercially usable reactor also play a substantial role as well as disposal requirements as far as they can be presently recognized. In the central chapter of this study, the documentation of the representatively selected concepts is evaluated as well as existing technology assessment studies and expert opinions. In a few cases where this appears to be necessary, according technical literature, further policy advisory reports, expert statements as well as other relevant sources are taken into account. Contradictions, different assessments and dissents in the literature as well as a few unsettled questions are thus indicated. The potential of advanced nuclear systems with respect to economical and societal as well as environmental objectives cannot exclusively be measured by the corresponding intrinsic or in comparison remarkable technical improvements. The acceptability of novel or improved systems in nuclear technology will have to be judged by their convincing solutions for the crucial questions of safety, nuclear waste and risk of proliferation of nuclear weapons

Neutron collar calibration for assay of LWR [light-water reactor] fuel assemblies

International Nuclear Information System (INIS)

Menlove, H.O.; Pieper, J.E.

1987-03-01

The neutron-coincidence collar is used for the verification of the uranium content in light-water reactor fuel assemblies. An AmLi neutron source is used to give an active interrogation of the fuel assembly to measure the 235 U content, and the 238 U content is verified from a passive neutron-coincidence measurement. This report gives the collar calibration data of pressurized-water reactor and boiling-water reactor fuel assemblies. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and different fuel assembly sizes. The data were collected at Exxon Nuclear, Franco-Belge de Fabrication de Combustibles, ASEA-Atom, and other nuclear fuel fabrication facilities

Experiences and Trends of Manufacturing Technology of Advanced Nuclear Fuels

International Nuclear Information System (INIS)

2012-08-01

The 'Atoms for Peace' mission initiated in the mid-1950s paved the way for the development and deployment of nuclear fission reactors as a source of heat energy for electricity generation in nuclear power reactors and as a source of neutrons in non-power reactors for research, materials irradiation, and testing and production of radioisotopes. The fuels for nuclear reactors are manufactured from natural uranium (∼99.3% 238 U + ∼0.7% 235 U) and natural thorium (∼100% 232 Th) resources. Currently, most power and research reactors use 235 U, the only fissile isotope found in nature, as fuel. The fertile isotopes 238 U and 232 Th are transmuted in the reactor to human-made 239 Pu and 233 U fissile isotopes, respectively. Likewise, minor actinides (MA) (Np, Am and Cm) and other plutonium isotopes are also formed by a series of neutron capture reactions with 238 U and 235 U. Long term sustainability of nuclear power will depend to a great extent on the efficient, safe and secure utilization of fissile and fertile materials. Light water reactors (LWRs) account for more than 82% of the operating reactors, followed by pressurized heavy water reactors (PHWRs), which constitute ∼10% of reactors. LWRs will continue to dominate the nuclear power market for several decades, as long as economically viable natural uranium resources are available. Currently, the plutonium obtained from spent nuclear fuel is subjected to mono recycling in LWRs as uranium-plutonium mixed oxide (MOX), containing up to 12% PuO 2 , in a very limited way. The reprocessed uranium (RepU) is also re-enriched and recycled in LWRs in a few countries. Unfortunately, the utilization of natural uranium resources in thermal neutron reactors is 2 and MOX fuel technology has matured during the past five decades. These fuels are now being manufactured, used and reprocessed on an industrial scale. Mixed uranium- plutonium monocarbide (MC), mononitride (MN) and U-Pu-Zr alloys are recognized as advanced fuels

Results of an aging-related failure survey of light water safety systems and components

International Nuclear Information System (INIS)

Meale, B.M.; Satterwhite, D.G.; MacDonald, P.E.

1988-01-01

The collection and evaluation of operating experience data are necessary in determining the effects of aging on the safety of operating nuclear plants. This paper presents the final results of a two-year research effort evaluating aging impacts on components in light water reactor systems. This research was performed as a part of the Nuclear Plant Aging Research program, sponsored by the US Nuclear Regulatory Commission. Two unique types of data analyses were performed. In the first, an aging-survey study, aging-related failure data for fifteen light water reactor systems were obtained from the Nuclear Plant Reliability Data System (NPRDS). These included safety, support, and power conversion systems. A computerized sort of these records classified each record into one of five generic categories, based on the utility's choice of the failure's NPRDS cause category. Systems and components within the systems that were most affected by aging were identified. In the second analysis, information on aging-related reported causes of failures was evaluated for component failures reported to NPRDS for auxiliary feedwater, high pressure injection, service water, and Class 1E electrical power distribution systems. 3 refs., 13 figs., 4 tabs

Advances in nuclear science and technology

CERN Document Server

Henley, Ernest J

1970-01-01

Advances in Nuclear Science and Technology, Volume 5 presents the underlying principles and theory, as well as the practical applications of the advances in the nuclear field. This book reviews the specialized applications to such fields as space propulsion.Organized into six chapters, this volume begins with an overview of the design and objective of the Fast Flux Test Facility to provide fast flux irradiation testing facilities. This text then examines the problem in the design of nuclear reactors, which is the analysis of the spatial and temporal behavior of the neutron and temperature dist

Advances in nuclear science and technology

CERN Document Server

Greebler, Paul

1966-01-01

Advances in Nuclear Science and Technology, Volume 3 provides an authoritative, complete, coherent, and critical review of the nuclear industry. This book presents the advances in the atomic energy field.Organized into six chapters, this volume begins with an overview of the use of pulsed neutron sources for the determination of the thermalization and diffusion properties of moderating as well as multiplying media. This text then examines the effect of nuclear radiation on electronic circuitry and its components. Other chapters consider radiation effects in various inorganic solids, with empha

Photoelectrochemical solar water splitting: From basic principles to advanced devices

Directory of Open Access Journals (Sweden)

Bandar Y.Alfaifi

2018-02-01

Full Text Available Photoelectrochemical water splitting (PEC offers a promising path for sustainable generation of hydrogen fuel. However, improving solar fuel water splitting efficiency facing tremendous challenges, due to the energy loss related to fast recombination of the photogenerated charge carriers, electrode degradation, as well as limited light harvesting. This review focuses on the brief introduction of basic fundamental of PEC water splitting and the concept of various types of water splitting approaches. Numerous engineering strategies for the investgating of the higher efficiency of the PEC, including charge separation, light harvesting, and co-catalysts doping, have been discussed. Moreover, recent remarkable progress and developments for PEC water splitting with some promising materials are discussed. Recent advanced applications of PEC are also reviewed. Finally, the review concludes with a summary and future outlook of this hot field.

An assessment of ex-vessel fuel-coolant interaction energetics for advanced light water reactors

International Nuclear Information System (INIS)

Murphy, J.G.; Corradini, M.L.

1997-01-01

The occurrence of an energetic fuel/coolant interaction (FCI) below the reactor pressure vessel in the cavity of advanced light water reactors (ALWRs) are analyzed to determine the possible hazard to structural walls as a result of dynamic liquid phase pressures. Such analyses are important to demonstrate that these cavity walls will maintain their integrity so that ex-vessel core debris coolability is possible. Past studies that have examined this or related issues are reviewed, and a methodology is proposed to analyze the occurrence of this physical event using the IFCI and TEXAS models for the FCI as well as dynamic shock wave propagation estimates using hand calculations as well as the CTH hydro model. Scenarios for the ALWRs are reviewed, and one severe accident scenario is used as an example to demonstrate the methodology. Such methodologies are recommended for consideration in future safety studies. These methodologies should be verified with direct comparison to energetic FCI data such as that being produced in KROTOS at the Joint Research Centre, Ispra

Effect of water chemistry improvement on flow accelerated corrosion in light-water nuclear reactor

International Nuclear Information System (INIS)

Sugino, Wataru; Ohira, Taku; Nagata, Nobuaki; Abe, Ayumi; Takiguchi, Hideki

2009-01-01

Flow Accelerated Corrosion (FAC) of Carbon Steel (CS) piping has been one of main issues in Light-Water Nuclear Reactor (LWRs). Wall thinning of CS piping due to FAC increases potential risk of pipe rupture and cost for inspection and replacement of damaged pipes. In particular, corrosion products generated by FAC of CS piping brought steam generator (SG) tube corrosion and degradation of thermal performance, when it intruded and accumulated in secondary side of PWR. To preserve SG integrity by suppressing the corrosion of CS, High-AVT chemistry (Feedwater pH9.8±0.2) has been adopted to Tsuruga-2 (1160 MWe PWR, commercial operation in 1987) in July 2005 instead of conventional Low-AVT chemistry (Feedwater pH 9.3). By the High-AVT adoption, the accumulation rate of iron in SG was reduced to one-quarter of that under conventional Low-AVT. As a result, a tendency to degradation of the SG thermal efficiency was improved. On the other hand, it was clarified that High-AVT is ineffective against Flow Accelerated Corrosion (FAC) at the region where the flow turbulence is much larger. By contrast, wall thinning of CS feed water pipes due to FAC has been successfully controlled by oxygen treatment (OT) for long time in BWRs. Because Magnetite film formed on CS surface under AVT chemistry has higher solubility and porosity in comparison with Hematite film, which is formed under OT. In this paper, behavior of the FAC under various pH and dissolved oxygen concentration are discussed based on the actual wall thinning rate of BWR and PWR plant and experimental results by FAC test-loop. And, it is clarified that the FAC is suppressed even under extremely low DO concentration such as 2ppb under AVT condition in PWR. Based on this result, we propose the oxygenated water chemistry (OWC) for PWR secondary system which can mitigate the FAC of CS piping without any adverse effect for the SG integrity. Furthermore, the applicability and effectiveness of this concept developed for FAC

Light Water Reactor Sustainability Program: Integrated Program Plan

Energy Technology Data Exchange (ETDEWEB)

NONE

2016-02-15

proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

Light Water Reactor Sustainability Program: Integrated Program Plan

International Nuclear Information System (INIS)

2016-02-01

and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

Light Water Reactor Sustainability Program: Integrated Program Plan

Energy Technology Data Exchange (ETDEWEB)

None, None

2017-05-01

proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

77 FR 13156 - Carolina Power & Light Company; Shearon Harris Nuclear Power Plant, Unit 1; Exemption

Science.gov (United States)

2012-03-05

... generation, and cladding oxidation from the metal/water reaction shall be calculated using the Baker-Just... boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within... spring 2012. The AREVA fuel design consists of low enriched uranium oxide fuel within M5 \\TM\\ zirconium...

Advanced Space Nuclear Reactors from Fiction to Reality

Science.gov (United States)

Popa-Simil, L.

The advanced nuclear power sources are used in a large variety of science fiction movies and novels, but their practical development is, still, in its early conceptual stages, some of the ideas being confirmed by collateral experiments. The novel reactor concept uses the direct conversion of nuclear energy into electricity, has electronic control of reactivity, being surrounded by a transmutation blanket and very thin shielding being small and light that at its very limit may be suitable to power an autonomously flying car. It also provides an improved fuel cycle producing minimal negative impact to environment. The key elements started to lose the fiction attributes, becoming viable actual concepts and goals for the developments to come, and on the possibility to achieve these objectives started to become more real because the theory shows that using the novel nano-technologies this novel reactor might be achievable in less than a century.

Hydrogen considerations in light-water power reactons

International Nuclear Information System (INIS)

Keilholtz, G.W.

1976-02-01

A critical review of the literature now available on hydrogen considerations in light-water power reactors (LWRs) and a bibliography of that literature are presented. The subject matter includes mechanisms for the generation of hydrogen-oxygen mixtures, a description of the fundamental properties of such mixtures, and their spontaneous ignition in both static and dynamic systems. The limits for hydrogen flammability and flame propagation are examined in terms of the effects of pressure, temperature, and additives; the emphasis is on the effects of steam and water vapor. The containment systems for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) are compared, and methods to control hydrogen and oxygen under the conditions of both normal operation and postulated accidents are reviewed. It is concluded that hydrogen can be controlled so that serious complications from the production of hydrogen will not occur. The bibliography contains abstracts from the computerized files of the Nuclear Safety Information Center. Key-word, author, and permuted-title indexes are provided. The bibliography includes responses to questions asked by the U. S. Nuclear Regulatory Commission (NRC) which relate to hydrogen, as well as information on normal operations and postulated accidents including generation of hydrogen from core sprays. Other topics included in the ten sections of the bibliography are metal-water reactions, containment atmosphere, radiolytic gas, and recombiners

Development and Deployment Strategy for a Small Advanced Light Water Reactor

International Nuclear Information System (INIS)

Modro, S. Michael; Reith, Raymond; Babka, Pierre

2002-01-01

This paper discusses development and deployment strategies for the modular Multi-Application Small Light Water Reactor (MASLWR). Modularity, small size, capability to transport whole modules including containment on road or by rail, simplicity and safety of this reactor allows innovative deployment strategies for a variety of applications. A larger plant may be constructed of many independent power generation units. The multi-module plant is intended to be operated as a base-load plant. Each reactor is to be operated at full load. However, in response to changes in power demand individual units can brought on line or shut down. A larger plant can be built in small increments to match the power demand balancing capital commitments with revenues from sales of electricity. Also, an unplanned shutdown of a reactor only affects a relatively small portion of the total plant capacity. Simplification of MASLWR design and extensive use of modularization coupled with factory fabrication will result in improved productivity of fieldwork and improved quality achieved in a factory environment. The initial MASLWR design concept development has been completed under the U.S. DOE (Department of Energy) Nuclear Energy Research Initiative (NERI) project. This paper discusses a strategy for developing and deploying a MASLWR plant by 2015. This schedule is realistic because the plant design relies on existing industrial experience and manufacturing capabilities. The development strategy consists of the following elements: concept confirmation through testing (under the NERI program a scaled integral test facility has been constructed and initial testing performed), design concept optimization, and design certification based on prototype testing. (authors)

Engineering for high heat loads on ALS [Advanced Light Source] beamlines

International Nuclear Information System (INIS)

DiGennaro, R.; Swain, T.

1989-08-01

This paper discussed general thermal engineering problems and specific categories of thermal design issues for high photon flux beam lines at the LBL Advanced Light Source: thermal distortion of optical surfaces and elevated temperatures of thermal absorbers receiving synchrotron radiation. A generic design for water-cooled heat absorbers is described for use with ALS photon shutters, beam defining apertures, and heat absorbing masks. Also, results of in- situ measurements of thermal distortion of a water-cooled mirror in a synchrotron radiation beam line are compared with calculated performance estimates. 17 refs., 2 figs

Effects of Water Radiolysis in Water Cooled Reactors - Nuclear Energy Research Initiative (NERI) Program

Energy Technology Data Exchange (ETDEWEB)

S. M. Pimblott

2000-10-01

OAK B188 Quarterly Progress Report on NERI Proposal No.99-0010 for the Development of an Experiment and Calculation Based Model to Describe the Effects of Radiation on Non-standard Aqueous Systems Like Those Encountered in the Advanced Light Water Reactor

Advances in Nuclear Power Plant Water Chemistry in Reducing Radiation Exposure

International Nuclear Information System (INIS)

Febrianto

2005-01-01

Water quality in light water reactor in Pressurized Water Reactor as well as in Boiling Water Reactor has being gradually improved since the beginning, to reduce corrosion risk and radiation exposure level. Corrosion problem which occurred to both type of reactors can reduce the plants availability, increase the operation and maintenance cost and increase the radiation exposure. Corrosion and radiation exposure risk in both reactor rare different. BWR type reactor has more experiences in corrosion problem because at the type of reactor lets water to boil in the core, while at PWR type reactor, water is kept not to boil. The BWR reactor has also higher radiation exposure rather than the PWR one. Many collaborative efforts of plants manufacturers and plant operator utilities have been done to reduce the radiation exposure level and corrosion risk. (author)

Fuel Summary Report: Shippingport Light Water Breeder Reactor

International Nuclear Information System (INIS)

Illum, D.B.; Olson, G.L.; McCardell, R.K.

1999-01-01

The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749

Regulatory Guide 1.131: Qualification tests of electric cables, field splices, and connections for light-water-cooled nuclear power plants

International Nuclear Information System (INIS)

Anon.

1992-01-01

Criterion III, ''Design Control,'' of Appendix B, ''Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plant,'' to 10 CFR Part 50, ''Licensing of Production and Utilization Facilities,'' requires that, where a test program is used to verify the adequacy of a specific design feature, it include suitable qualification testing of a prototype unit under the most adverse design conditions. This regulatory guide describes a method acceptable to the NRC staff for complying with the Commission's regulations with regard to qualification testing of electric cables, field splices, and connections for service in light-water-cooled nuclear power plants to ensure that the cables, field splices, and connections can perform their safety-related functions. The fire test provisions of this guide do not apply to qualification for an installed configuration

Taking a fresh Look at boiling heat transfer on the road to improved nuclear economics and efficiency

Energy Technology Data Exchange (ETDEWEB)

Baglietto, E.; Pointer, W. D.

2016-08-01

In the effort to reinvigorate innovation in the way we design, build, and operate the nuclear power generating stations of today and tomorrow, nothing can be taken for granted. Not even the seemingly familiar physics of boiling water. The Consortium for the Advanced Simulation of Light Water Reactors, or CASL, is focused on the deployment of advanced modeling and simulation capabilities to enable the nuclear industry to reduce uncertainties in the prediction of multi-physics phenomena and continue to improve the performance of todays light water reactors and their fuel. An important part of the CASL mission is the development of a next generation thermal hydraulics simulation capability, integrating the history of engineering models based on based on experimental experience with the computing technology of the future. (Author)

Is light water reactor technology sustainable?

International Nuclear Information System (INIS)

Rothwell, G.; Van der Zwaan, B.

2001-01-01

This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

ULTRA SCWR+: Practical advanced water reactor concepts

International Nuclear Information System (INIS)

Duffey, Romney; Khartabil, Hussam; Kuran, Sermet; Zhou, Tracy; Pioro, Igor

2008-01-01

Modern thermal power plants now utilize supercritical steam cycles with thermal efficiencies of over 45%. Recent developments have lead to Ultra-SuperCritical (USC) systems, which adopt reheat turbines that can attain efficiencies of over 50%. Because these turbines are already developed, demonstrated and deployed worldwide, and use existing and traditional steam cycle technology, the simplest nuclear advance is to utilize these proven thermal cycle conditions by coupling this turbine type to a reactor. This development direction is fundamentally counter to the usual approach of adopting high-temperature gas-cooled (helium-cooled) reactor cycles, for which turbines have yet to be demonstrated on commercial scale unlike the supercritical steam turbines. The ULTRA (Ultra-supercritical Light water Thermal ReActor) SCWR+ concept adopts the fundamental design approach of matching a water and steam-cooled reactor to the ultra-supercritical steam cycle, adopting the existing and planned thermal power plant turbines. The HP and IP sections are fed with conditions of 25 MPa/625degC and 7 MPa/700degC, respectively, to achieve operating plant thermal efficiencies in excess of 50%, with a direct turbine cycle. By using such low-pressure reheated steam, this concept also adopts technology that was explored and used many years ago in existing water reactors, with the potential to produce large quantities of low cost heat, which can be used for other industrial and district processes. Pressure-Tube (PT) reactors are suitable for adoption of this design approach and, in addition, have other advantages that will significantly improve water-cooled reactor technology. These additional advantages include enhanced safety and improved resource utilization and proliferation resistance. This paper describes the PT-SCWR+ concept and its potential enhancements. (author)

Advanced control and instrumentation systems in nuclear power plants. Design, verification and validation

International Nuclear Information System (INIS)

Haapanen, P.

1995-01-01

The Technical Committee Meeting on design, verification and validation of advanced control and instrumentation systems in nuclear power plants was held in Espoo, Finland on 20 - 23 June 1994. The meeting was organized by the International Atomic Energy Agency's (IAEA) International Working Group's (IWG) on Nuclear Power Plant Control and Instrumentation (NPPCI) and on Advanced Technologies for Water Cooled Reactors (ATWR). VTT Automation together with Imatran Voima Oy and Teollisuuden Voima Oy responded about the practical arrangements of the meeting. In total 96 participants from 21 countries and the Agency took part in the meeting and 34 full papers and 8 posters were presented. Following topics were covered in the papers: (1) experience with advanced and digital systems, (2) safety and reliability analysis, (3) advanced digital systems under development and implementation, (4) verification and validation methods and practices, (5) future development trends. (orig.)

Light-water reactor research and development

International Nuclear Information System (INIS)

1985-05-01

This report on the national program of research and development on light water reactors is the second of two reports requested in 1982 by W. Kenneth Davis, Deputy Secretary of the Department of Energy. A first report, published in September 1983, treated the needs for safety-related R and D. In this second report, the Energy Research Advisory Board finds that, although many light water reactors are providing reliable and economic electricity, it appears unlikely that U.S. utilities will order additional reactors until the currently unacceptable economic risk, created by the regulatory climate and uncertain demand, is reduced. Thus it is unlikely that the private sector alone will fund major LWR design improvements. However, nuclear power will continue on its current course of expansion overseas. DOE participation is vitally needed to support the national interest in LWR technology. The report outlines R and D needs for a program to improve the safety, reliability, and economics of the present generation of plants; to develop evolutionary improved designs to be ready when needed; and to explore innovative longer-term concepts for deployment after the year 2000. The respective roles of government and the private sector are discussed

Towards intrinsically safe light-water reactors

International Nuclear Information System (INIS)

Hannerz, K.

1983-02-01

The reactor-safety issue is one of the principal problems threatening the future of the nuclear option, at least in participatory democracies. It has contributed to widespread public distrust and is the direct cause of the escalation in design complexity and quality assurance requirements that are rapidly eroding the competitive advantage of nuclear power. Redesign of the light-water reactor can eliminate those features that leave it open to public distrust and obstructive intervention. This redesign appears feasible within the realm of proven technology in those fields (fuels, materials, water chemistry, waste technology, etc.) in which extended operating experience is essential for confidence in system performance. A pressurized water reactor outline design developed to achieve the above goal is presented. The key feature is the design of the primary system extracting heat from the core so that the latter is protected from damage caused by any credible system failure or any destructive intervention from the outside by either violent means (up to and including nonnuclear warfare) or by mistaken or malicious use of the plant control systems. Such a design objective can be achieved by placing the entire primary circulation system in a large pressurized pool of cold water with a high boric acid content. Enough water is provided in the pool to allow core-decay-heat removal by evaporation for at least one week following any incident with no cooling systems operating. Subsequently it is assumed that a supply of further water (a few cubic meters per hour) from the outside can be arranged, even without the presence of the plant operating personnel

Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: reprocessing light-water reactor fuel

International Nuclear Information System (INIS)

Finney, B.C.; Blanco, R.E.; Dahlman, R.C.; Hill, G.S.; Kitts, F.G.; Moore, R.E.; Witherspoon, J.P.

1976-10-01

A cost/benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model nuclear fuel reprocessing plant which processes light-water reactor (LWR) fuels, and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term as low as reasonably achievable in relation to limiting the release of radioactive materials from nuclear facilities. The base case model plant is representative of current plant technology and has an annual capacity of 1500 metric tons of LWR fuel. Additional radwaste treatment systems are added to the base case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The cost for the added waste treatment operations and the corresponding dose commitments are calculated for each case. In the final analysis, radiological dose is plotted vs the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Much of the technology used in the advanced cases is in an early stage of development and is not suitable for immediate use. The methodology used in estimating the costs, and the radiological doses, detailed calculations, and tabulations are presented in Appendix A and ORNL-4992. This report is a revision of the original study

Functional issues and environmental qualification of digital protection systems of advanced light-water nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Korsah, K.; Clark, R.L.; Wood, R.T. [Oak Ridge National Lab., TN (United States)

1994-04-01

Issues of obsolescence and lack of infrastructural support in (analog) spare parts, coupled with the potential benefits of digital systems, are driving the nuclear industry to retrofit analog instrumentation and control (I&C) systems with digital and microprocessor-based systems. While these technologies have several advantages, their application to safety-related systems in nuclear power plants raises key issues relating to the systems` environmental qualification and functional reliability. To bound the problem of new I&C system functionality and qualification, the authors focused this study on protection systems proposed for use in ALWRs. Specifically, both functional and environmental qualification issues for ALWR protection system I&C were addressed by developing an environmental, functional, and aging data template for a protection division of each proposed ALWR design. By using information provided by manufacturers, environmental conditions and stressors to which I&C equipment in reactor protection divisions may be subjected were identified. The resulting data were then compared to a similar template for an instrument string typically found in an analog protection division of a present-day nuclear power plant. The authors also identified fiber-optic transmission systems as technologies that are relatively new to the nuclear power plant environment and examined the failure modes and age-related degradation mechanisms of fiber-optic components and systems. One reason for the exercise of caution in the introduction of software into safety-critical systems is the potential for common-cause failure due to the software. This study, however, approaches the functionality problem from a systems point of view. System malfunction scenarios are postulated to illustrate the fact that, when dealing with the performance of the overall integrated system, the real issues are functionality and fault tolerance, not hardware vs. software.

Functional issues and environmental qualification of digital protection systems of advanced light-water nuclear reactors

International Nuclear Information System (INIS)

Korsah, K.; Clark, R.L.; Wood, R.T.

1994-04-01

Issues of obsolescence and lack of infrastructural support in (analog) spare parts, coupled with the potential benefits of digital systems, are driving the nuclear industry to retrofit analog instrumentation and control (I ampersand C) systems with digital and microprocessor-based systems. While these technologies have several advantages, their application to safety-related systems in nuclear power plants raises key issues relating to the systems' environmental qualification and functional reliability. To bound the problem of new I ampersand C system functionality and qualification, the authors focused this study on protection systems proposed for use in ALWRs. Specifically, both functional and environmental qualification issues for ALWR protection system I ampersand C were addressed by developing an environmental, functional, and aging data template for a protection division of each proposed ALWR design. By using information provided by manufacturers, environmental conditions and stressors to which I ampersand C equipment in reactor protection divisions may be subjected were identified. The resulting data were then compared to a similar template for an instrument string typically found in an analog protection division of a present-day nuclear power plant. The authors also identified fiber-optic transmission systems as technologies that are relatively new to the nuclear power plant environment and examined the failure modes and age-related degradation mechanisms of fiber-optic components and systems. One reason for the exercise of caution in the introduction of software into safety-critical systems is the potential for common-cause failure due to the software. This study, however, approaches the functionality problem from a systems point of view. System malfunction scenarios are postulated to illustrate the fact that, when dealing with the performance of the overall integrated system, the real issues are functionality and fault tolerance, not hardware vs. software

Nuclide inventories of spent fuels from light water reactors

International Nuclear Information System (INIS)

Okumura, Keisuke; Okamoto, Tsutomu

2012-02-01

Accurate information on nuclide inventories of spent fuels from Light Water Reactors (LWRs) is important for evaluations of criticality, decay heat, radioactivity, toxicity, and so on, in the safety assessments of storage, transportation, reprocessing and waste disposal of the spent fuels. So, a lot of lattice burn-up calculations were carried out for the possible fuel specifications and irradiation conditions in Japanese commercial LWRs by using the latest nuclear data library JENDL-4.0 and a sophisticated lattice burn-up calculation code MOSRA-SRAC. As a result, burn-up changes of nuclide inventories and their possible ranges were clarified for 21 heavy nuclides and 118 fission products, which are important from the viewpoint of impacts to nuclear characteristics and nuclear fuel cycle and environment. (author)

Waste disposal from the light water reactor fuel cycle

International Nuclear Information System (INIS)

Costello, J.M.; Hardy, C.J.

1981-05-01

Alternative nuclear fuel cycles for support of light water reactors are described and wastes containing naturally occurring or artificially produced radioactivity reviewed. General principles and objectives in radioactive waste management are outlined, and methods for their practical application to fuel cycle wastes discussed. The paper concentrates upon management of wastes from upgrading processes of uranium hexafluoride manufacture and uranium enrichment, and, to a lesser extent, nuclear power reactor wastes. Some estimates of radiological dose commitments and health effects from nuclear power and fuel cycle wastes have been made for US conditions. These indicate that the major part of the radiological dose arises from uranium mining and milling, operation of nuclear reactors, and spent fuel reprocessing. However, the total dose from the fuel cycle is estimated to be only a small fraction of that from natural background radiation

Technical characteristics of new generation of nuclear power plants; Charakterystyka techniczna elektrowni jadrowych nowej generacji

Energy Technology Data Exchange (ETDEWEB)

Janczak, R.; Mikulski, A.; Staron, E. [Instytut Energii Atomowej, Swierk-Otwock (Poland)

1997-12-31

The concept of Advanced Light Water Reactors (ALWR) as a new generation of nuclear reactors for energetics have been presented. The influence of reactor accidents (TMI and Chernobyl) on technical and scientific development of nuclear reactors has been discussed from the view point of safety assurance and requirements being defined by American and European Nuclear Regulatory commission. 12 refs, 14 figs.

Method of neutronic calculations for a spherical cell equivalent to cylindrical one for using computer codes in light water reactors in the fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Borges, V.; Sefidvash, F.; Rastogi, E.P.; Huria, H.C.; Krishnani, P.D.

1989-01-01

In order to use the existing light water reactor cell calculation codes for fluidized bed nuclear reactor having spherical fuel cells, an equivalence method has been developed. This method is shown to be adequate in calculation of the Dancoff factor. This method also was applicable in LEOPARD code and the results obtained in calculation of K ∞ was compared with the obtained using the DTF IV code, the results showed that the method is adequate for the calculations neutronics of the fluidized bed nuclear reactor. (author) [pt

Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-04-30

The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.

Implications of alpha-decay for long term storage of advanced heavy water reactor fuels

International Nuclear Information System (INIS)

Pencer, J.; McDonald, M.H.; Roubtsov, D.; Edwards, G.W.R.

2017-01-01

Highlights: •Alpha decays versus storage time are calculated for examples of advanced heavy water reactor fuels. •Estimates are made for fuel swelling and helium bubble formation as a function of time. •These predictions are compared to predictions for natural uranium fuel. •Higher rates of damage are predicted for advanced heavy water reactor fuels than natural uranium. -- Abstract: The decay of actinides such as 238 Pu, results in recoil damage and helium production in spent nuclear fuels. The extent of the damage depends on storage time and spent fuel composition and has implications for the integrity of the fuels. Some advanced nuclear fuels intended for use in pressurized heavy water pressure tube reactors have high initial plutonium content and are anticipated to exhibit swelling and embrittlement, and to accumulate helium bubbles over storage times as short as hundreds of years. Calculations are performed to provide estimates of helium production and fuel swelling associated with alpha decay as a function of storage time. Significant differences are observed between predicted aging characteristics of natural uranium and the advanced fuels, including increased helium concentrations and accelerated fuel swelling in the latter. Implications of these observations for long term storage of advanced fuels are discussed.

Water chemistry diagnosis system for nuclear power plants

International Nuclear Information System (INIS)

Igarashi, Hiroo; Koya, Hiroshi; Osumi, Katsumi.

1990-01-01

The water quality control for the BWRs in Japan has advanced rapidly recently, and as to the dose reduction due to the decrease of radioactivity, Japan takes the position leading the world. In the background of the advanced water quality control like this and the increase of nuclear power plants in operation, the automation of arranging a large quantity of water quality control information and the heightening of its reliability have been demanded. Hitachi group developed the water quality synthetic control system which comprises the water quality data management system to process a large quantity of water quality data with a computer and the water quality diagnosis system to evaluate the state of operation of the plants by the minute change of water quality and to carry out the operational guide in the aspect of water quality control. To this water quality diagnosis system, high speed fuzzy inference is applied in order to do rapid diagnosis with fuzzy data. The trend of development of water quality control system, the construction of the water quality synthetic control system, the configuration of the water quality diagnosis system and the development of algorithm and the improvement of the reliability of maintenance are reported. (K.I.)

Design and analysis challenges for advanced nuclear fuel

International Nuclear Information System (INIS)

Klepfer, H.; Abdollahian, D.; Dias, A.; Durston, C.; Eisenhart, L.; Engel, R.; Gilmore, P.; Rank, P.; Kjaer-Pedersen, N.; Sorensen, J.; Yang, R.; Agee, L.

2004-01-01

Significant changes have been incorporated in the light water reactor (LWR) fuel designs now being offered, and advanced fuel designs are currently being developed for the existing and the next generation of reactor designs. These advanced fuel design configurations are intended to offer utilities major economic gains, including: (1) improved fuel characteristics through optimized hydrogen to uranium ratio within the core; (2) increased capacity factor by allowing longer operating cycles, which is implemented by increasing the fuel enrichment and the amount and distribution of burnable poison, gadolinia, boron, or erbium within the fuel assembly to achieve higher discharge burnup; and (3) increased plant power output, if it can be accommodated by the balance of plant, by increasing the power density of the fuel assembly. The authors report here work being done to identify emerging technical issues in support of utility industry evaluations of advanced fuel designs. (author)

Generic safety issues for nuclear power plants with light water reactors and measures taken for their resolution

International Nuclear Information System (INIS)

1998-09-01

The IAEA Conference on 'The Safety of Nuclear Power: Strategy for the Future' in 1991 was a milestone in nuclear safety. Two of the important items addressed by this conference were ensuring and enhancing safety of operating plants and treatment of nuclear power plants built to earlier safety standards. A number of publications related to these two items issued subsequent to this conference were: A Common Basis for Judging the Safety of Nuclear Power Plants Built to Earlier Standards, INSAG-9 (1995), the IAEA Safety Guide 50-SG-O12, periodic Safety Review of Operational Nuclear Power Plants (1994) and an IAEA publication on the Safety Evaluation of Operating Nuclear Power Plants Built to Earlier Standards - A Common Basis for Judgement (1997). Some of the findings of the 1991 Conference have not yet been fully addressed. An IAEA Symposium on reviewing the Safety of Existing Nuclear Power Plants in 1996 showed that there is an urgent need for operating organizations and national authorities to review operating nuclear power plants which do not meet the high safety levels of the vast majority of plants and to undertake improvements with assistance from the international community if required. Safety reviews of operating nuclear power plants take on added importance in the context of the Convention on Nuclear safety and its implementation. The purpose of this TECDOC compilation based on broad international experience, is to assist the Member States in the reassessment of operating plants by providing a list of generic safety issues identified in nuclear power plants together with measures taken to resolve these issues. These safety issues are generic in nature with regard to light water reactors and the measures for their resolution are for use as a reference for the safety reassessment of operating plants. The TECDOC covers issues thought to be significant to Member States based on consensus process. It provides an introduction to the use of generic safety issues for

Nuclear structure studies with low-energy light ions: fundamental and applied

International Nuclear Information System (INIS)

Mazumdar, I.

2016-01-01

Studies in low and medium energy nuclear physics have been dominated by heavy-ion induced reactions for last five decades. Heavy-ion induced nuclear reactions have enriched our knowledge of the structural evolutions and intricacies of reaction dynamics of the nuclear many-body systems. However, the emergence and rise of heavy-ion physics have seen a general decline in studies with low- and medium-energy light-ion beams. The harsh reality of dwindling number of low-energy light ion facilities adversely affect research in nuclear physics. Very low-energy and high current light-ion facilities immediately conjures up in our minds the studies in nuclear astrophysics. Measurements of light-ion capture cross sections and astrophysical S factors are the major themes of research at most of the light-ion facilities. However, the importance low energy light-ion beams is multifarious. A variety of measurements providing vital support and inputs to heavy-ion research can only be carried out at the low-energy, light-ion facilities. Light-ion beams are also useful for generation of mono-energetic neutron beams. In this talk I will draw from some of our recent measurements to show the importance of light-ion beams in nuclear astrophysics and also in applied nuclear physics. (author)

Development trends in light water reactors

International Nuclear Information System (INIS)

Fogelstroem, L.; Simon, M.

1988-01-01

The present market for new nuclear power plants is weak, but is expected to pick up again, which is why great efforts are being made to further develop the light water reactor line for future applications. There is both a potential and a need for further improvement, for instance with respect to even higher cost efficiency, a simplified operating permit procedure, shorter construction periods, and increased operational flexibility to meet rising demands in load following behavior and in better cycle data of fuel elements. However, also public acceptance must not be forgotten when deciding about the line to be followed in the development of LWR technology. (orig.) [de

Is light water reactor technology sustainable?

Energy Technology Data Exchange (ETDEWEB)

Rothwell, G. [Stanford Univ., Dept. of Economics, CA (United States); Van der Zwaan, B. [Vrije Univ., Amsterdam, Inst. for Environmental Studies (Netherlands)

2001-07-01

This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

Advances in nuclear fuel technology. 3. Development of advanced nuclear fuel recycle systems

International Nuclear Information System (INIS)

Arie, Kazuo; Abe, Tomoyuki; Arai, Yasuo

2002-01-01

Fast breeder reactor (FBR) cycle technology has a technical characteristics flexibly easy to apply to diverse fuel compositions such as plutonium, minor actinides, and so on and fuel configurations. By using this characteristics, various feasibilities on effective application of uranium resources based on breeding of uranium of plutonium for original mission of FBR, contribution to radioactive wastes problems based on amounts reduction of transuranium elements (TRU) in high level radioactive wastes, upgrading of nuclear diffusion resistance, extremely upgrading of economical efficiency, and so on. In this paper, were introduced from these viewpoints, on practice strategy survey study on FBR cycle performed by cooperation of the Japan Nuclear Cycle Development Institute (JNC) with electric business companies and so on, and on technical development on advanced nuclear fuel recycle systems carried out at the Central Research Institute of Electric Power Industry, Japan Atomic Energy Research Institute, and so on. Here were explained under a vision on new type of fuels such as nitride fuels, metal fuels, and so on as well as oxide fuels, a new recycle system making possible to use actinides except uranium and plutonium, an 'advanced nuclear fuel cycle technology', containing improvement of conventional wet Purex method reprocessing technology, fuel manufacturing technology, and so on. (G.K.)

New nuclear plant design and licensing process

International Nuclear Information System (INIS)

Luangdilok, W.

1996-01-01

This paper describes latest developments in the nuclear power reactor technology with emphasis on three areas: (1) the US technology of advanced passive light water reactors (AP600 and S BWR), (2) regulatory processes that certify their safety, and (3) current engineering concerns. The goal is to provide and insight of how the government's regulatory agency guarantees public safety by looking into how new passive safety features were designed and tested by vendors and how they were re-evaluated and retested by the US NRC. The paper then discusses the US 1989 nuclear licensing reform (10 CFR Part 52) whose objectives are to promote the standardization of nuclear power plants and provide for the early and definitive resolution of site and design issues before plants are built. The new licensing process avoids the unpredictability nd escalated construction cost under the old licensing process. Finally, the paper summarizes engineering concerns found in current light water reactors that may not go away in the new design. The concerns are related the material and water chemistry technology in dealing with corrosion problems in water-cooled nuclear reactor systems (PWRs and BWRs). These engineering concerns include core shroud cracking (BWRs), jet pump hold-down beam cracking (BWRs), steam generator tube stress corrosion cracking (PWR)

Hydrogen production from water: Recent advances in photosynthesis research

Energy Technology Data Exchange (ETDEWEB)

Greenbaum, E.; Lee, J.W. [Oak Ridge National Lab., TN (United States). Chemical Technology Div.

1997-12-31

The great potential of hydrogen production by microalgal water splitting is predicated on quantitative measurement of the algae`s hydrogen-producing capability, which is based on the following: (1) the photosynthetic unit size of hydrogen production; (2) the turnover time of photosynthetic hydrogen production; (3) thermodynamic efficiencies of conversion of light energy into the Gibbs free energy of molecular hydrogen; (4) photosynthetic hydrogen production from sea water using marine algae; (5) the potential for research advances using modern methods of molecular biology and genetic engineering to maximize hydrogen production. ORNL has shown that sustained simultaneous photoevolution of molecular hydrogen and oxygen can be performed with mutants of the green alga Chlamydomonas reinhardtii that lack a detectable level of the Photosystem I light reaction. This result is surprising in view of the standard two-light reaction model of photosynthesis and has interesting scientific and technological implications. This ORNL discovery also has potentially important implications for maximum thermodynamic conversion efficiency of light energy into chemical energy by green plant photosynthesis. Hydrogen production performed by a single light reaction, as opposed to two, implies a doubling of the theoretically maximum thermodynamic conversion efficiency from {approx}10% to {approx}20%.

Estabilishing requirements for the next generation of pressurized water reactors--reducing the uncertainty

International Nuclear Information System (INIS)

Chernock, W.P.; Corcoran, W.R.; Rasin, W.H.; Stahlkopf, K.E.

1987-01-01

The Electric Power Research Institute is managing a major effort to establish requirements for the next generation of U.S. light water reactors. This effort is the vital first step in preserving the viability of the nuclear option to contribute to meet U.S. national electric power capacity needs in the next century. Combustion Engineering, Inc. and Duke Power Company formed a team to participate in the EPRI program which is guided by a Utility Steering committee consisting of experienced utility technical executives. A major thrust of the program is to reduce the uncertainties which would be faced by the utility executives in choosing the nuclear option. The uncertainties to be reduced include those related to safety, economic, operational, and regulatory aspects of advanced light water reactors. This paper overviews the Requirements Document program as it relates to the U.S. Advanced Light Water Reactor (ALWR) effort in reducing these uncertainties and reports the status of efforts to establish requirements for the next generation of pressurized water reactors. It concentrates on progress made in reducing the uncertainties which would deter selection of the nuclear option for contributing to U.S. national electric power capacity needs in the next century and updates previous reports in the same area. (author)

Flow-induced vibration for light water reactors. Progress report, January-June 1980

International Nuclear Information System (INIS)

De Coster, M.A.

1981-02-01

Flow-Induced Vibration for Light Water Reactors (FIV for LWRs) is a four-year program designed to improve the FIV performance of light water reactors through the development of design criteria, analytical models for predicting behavior of components, general scaling laws to improve the accuracy of reduced-scale tests, and the identification of high FIV risk areas. The program is managed by the General Electric Nuclear Power Systems Engineering Department and has three major contributors: General Electric Nuclear Power Systems Engineering Department (NPSED), General Electric Corporate Research and Development (CR and D) and Argonne National Laboratory (ANL). The program commenced December 1, 1976, but was suspended on September 30, 1978, due to a shift in Department of Energy (DOE) priorities away from LWR productivity/availability. It was reinitiated as of August 1, 1979. A second program suspension occurred from March 29, 1980 through May 16, 1980, due to funding limits. This progress report summarizes the accomplishments achieved during the period from Janury 1980 to June 1980

Practical considerations in the use of UV light for drinking water disinfection

International Nuclear Information System (INIS)

Jeyanayagam, S.; Cotton, C.

2002-01-01

Ultraviolet (UV) light was discovered approximately 150 years ago. The first commercial UV lamp was made in the early 1900s soon followed by the manufacture of the quartz sleeve. These technological advances allowed the first application of UV light for water disinfection in 1907 in France. In the mid 1980s, UV disinfection was named as a Best available technology (BAT) for wastewater disinfection in the United States. Fueled by the recent findings that UV disinfection can inactivate key pathogens at cost effective UV doses, the drinking water industry in North America is closely investigating its application in large installations. (author)

Study of Pu consumption in Advanced Light Water Reactors

International Nuclear Information System (INIS)

1993-01-01

Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE's 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology

Enhancing proliferation resistance in advanced light water reactor fuel cycles

International Nuclear Information System (INIS)

Kazimi, M.S.; Pilat, E.E.; Driscoll, M.J.; Xu, Z.; Wang, D.; Zhao, X.

2001-01-01

Alternative once-through, light water reactor fuel designs are evaluated for capability to reduce the amount and quality of plutonium produced. Doubling the discharge burnup is quite effective, producing modest reductions in total plutonium and significant increases in 238 Pu whose heat generation and spontaneous neutrons complicate weapon usability. Reductions in the hydrogen to heavy metal ratio are counterproductive. Increases are helpful, but only small changes can be accommodated. Use of ThO 2 in a homogeneous mixture with UO 2 can reduce plutonium production to about 50% of that in a typical present day PWR, and in heterogeneous seed-blanket designs can reduce it to 30 to 45%. (author)

RISMC advanced safety analysis project plan: FY2015 - FY2019. Light Water Reactor Sustainability Program

International Nuclear Information System (INIS)

Szilard, Ronaldo H; Smith, Curtis L; Youngblood, Robert

2014-01-01

In this report, the Advanced Safety Analysis Program (ASAP) objectives and value proposition is described. ASAP focuses on modernization of nuclear power safety analysis (tools, methods and data); implementing state-of-the-art modeling techniques (which include, for example, enabling incorporation of more detailed physics as they become available); taking advantage of modern computing hardware; and combining probabilistic and mechanistic analyses to enable a risk informed safety analysis process. The modernized tools will maintain the current high level of safety in our nuclear power plant fleet, while providing an improved understanding of safety margins and the critical parameters that affect them. Thus, the set of tools will provide information to inform decisions on plant modifications, refurbishments, and surveillance programs, while improving economics. The set of tools will also benefit the design of new reactors, enhancing safety per unit cost of a nuclear plant. As part of the discussion, we have identified three sets of stakeholders, the nuclear industry, the Department of Energy (DOE), and associated oversight organizations. These three groups would benefit from ASAP in different ways. For example, within the DOE complex, the possible applications that are seen include the safety of experimental reactors, facility life extension, safety-by-design in future generation advanced reactors, and managing security for the storage of nuclear material. This report provides information in five areas: (1) A value proposition (@@@why is this important?@@@) that will make the case for stakeholder's use of the ASAP research and development (R&D) products; (2) An identification of likely end users and pathway to adoption of enhanced tools by the end-users; (3) A proposed set of practical and achievable @@use case@@@ demonstrations; (4) A proposed plan to address ASAP verification and validation (V&V) needs; and (5) A proposed schedule for the multi-year ASAP.

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

International Nuclear Information System (INIS)

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

Energy Technology Data Exchange (ETDEWEB)

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

International Nuclear Information System (INIS)

Ragusa, Jean; Vierow, Karen

2011-01-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

Energy Technology Data Exchange (ETDEWEB)

Ragusa, Jean; Vierow, Karen

2011-09-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

Directory of Open Access Journals (Sweden)

Kulesza Joel A.

2016-01-01

Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

Non-chemical water purification a Westinghouse/Wallenius product for nuclear power plant needs

International Nuclear Information System (INIS)

Goetberg, J.; Carlsson, M.

2014-01-01

Increasing demand for ecologically effective water treatment technologies has resulted in the development of several new oxidation methods. These technologies are generally labelled Advanced Oxidation Technologies (AOT) or Advanced Oxidation Processes (AOP) and currently represent the most widely recognized alternative for ecologically sound, high-tech water purification. Many years of intensive research have culminated in the innovative Wallenius-AOT technology, a patented method that is remarkable in several ways. It imitates nature's own water purification method. This means no chemical additives are needed. The technology utilizes the ability of light, together with photo-catalytic semiconductor surfaces, to produce free radicals, like nature does. These reactive radicals create an environment in which organic and inorganic substances oxidize, whereby a broad spectrum of organisms is rendered harmless more effectively than with conventional UV technology. The entire process takes just a few micro-seconds. A major advantage of the technology is that it can be adjusted according to the desired degree of purification. By altering the dynamics of the process, the purification can be designed for specific applications. In this way, AOT tackles precise problems, regardless of flow and whether the problem is chemical or biological. The product was originally introduced for ballast treatment in the shipping industry. Ballast water has created severe damages to the biology at many locations. By moving an organism from one ocean to another we have introduced a possible threat to the local ecosystem. This has been prevented by using the AOT water treatment units. During ballasting and de-ballasting, the units create radicals with the help of a catalyst and a light source. These radicals then destroy the cell membrane of microorganisms. The radicals, which never leave the unit, have a lifetime of only a few milliseconds and pose no risk to the environment or crew

Nuclear generating station and heavy water plant cost estimates for strategy studies

International Nuclear Information System (INIS)

Archinoff, G.H.

1979-07-01

Nuclear generating station capital, operating and maintenance costs are basic input data for strategy analyses of alternate nuclear fuel cycles. This report presents estimates of these costs for natural uranium CANDU stations, CANDU stations operating on advanced fuel cycles, and liquid metal fast breeder reactors. Cost estimates for heavy water plants are also presented. The results show that station capital costs for advanced fuel cycles are not expected to be significantly greater than those for natural uranium stations. LMFBR capital costs are expected to be 25-30 percent greater than for CANDU's. (auth)

Hydrogen behavior in light-water reactors

International Nuclear Information System (INIS)

Berman, M.; Cummings, J.C.

1984-01-01

The Three Mile Island accident resulted in the generation of an estimated 150 to 600 kg of hydrogen, some of which burned inside the containment building, causing a transient pressure rise of roughly 200 kPa (2 atm). With this accident as the immediate impetus and the improved safety of reactors as the long-term goal, the nuclear industry and the Nuclear Regulatory Commission initiated research programs to study hydrogen behavior and control during accidents at nuclear plants. Several fundamental questions and issues arise when the hydrogen problem for light-water-reactor plants is examined. These relate to four aspects of the problem: hydrogen production; hydrogen transport, release, and mixing; hydrogen combustion; and prevention or mitigation of hydrogen combustion. Although much has been accomplished, some unknowns and uncertainties still remain, for example, the rate of hydrogen production during a degraded-core or molten-core accident, the rate of hydrogen mixing, the effect of geometrical structures and scale on combustion, flame speeds, combustion completeness, and mitigation-scheme effectiveness. This article discusses the nature and extent of the hydrogen problem, the progress that has been made, and the important unresolved questions

Nuclear Physics on the Light Front

OpenAIRE

Miller, Gerald A.

1999-01-01

High energy scattering experiments involving nuclei are typically analyzed in terms of light front variables. The desire to provide realistic, relativistic wave functions expressed in terms of these variables led me to try to use light front dynamics to compute nuclear wave functions. The progress is summarized here.

ATLAS program for advanced thermal-hydraulic safety research

International Nuclear Information System (INIS)

Song, Chul-Hwa; Choi, Ki-Yong; Kang, Kyoung-Ho

2015-01-01

Highlights: • Major achievements of the ATLAS program are highlighted in conjunction with both developing advanced light water reactor technologies and enhancing the nuclear safety. • The ATLAS data was shown to be useful for the development and licensing of new reactors and safety analysis codes, and also for nuclear safety enhancement through domestic and international cooperative programs. • A future plan for the ATLAS testing is introduced, covering recently emerging safety issues and some generic thermal-hydraulic concerns. - Abstract: This paper highlights the major achievements of the ATLAS program, which is an integral effect test program for both developing advanced light water reactor technologies and contributing to enhancing nuclear safety. The ATLAS program is closely related with the development of the APR1400 and APR"+ reactors, and the SPACE code, which is a best-estimate system-scale code for a safety analysis of nuclear reactors. The multiple roles of ATLAS testing are emphasized in very close conjunction with the development, licensing, and commercial deployment of these reactors and their safety analysis codes. The role of ATLAS for nuclear safety enhancement is also introduced by taking some examples of its contributions to voluntarily lead to multi-body cooperative programs such as domestic and international standard problems. Finally, a future plan for the utilization of ATLAS testing is introduced, which aims at tackling recently emerging safety issues such as a prolonged station blackout accident and medium-size break LOCA, and some generic thermal-hydraulic concerns as to how to figure out multi-dimensional phenomena and the scaling issue.

Nuclear Reactor Engineering Analysis Laboratory

International Nuclear Information System (INIS)

Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

1998-01-01

The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels

Advanced Reactor Development in the United States

Energy Technology Data Exchange (ETDEWEB)

Giessing, D. F.; Griffith, J. D.; McGoff, D. J.; Rosen, Sol [U. S. Department of Energy, Texas (United States)

1990-04-15

In the United States, three technologies are employed for the new generation of advanced reactors. These technologies are Advanced Light Water Reactors (A LWRs) for the 1990s and beyond, the Modular High Temperature Gas Reactor (M HTGR) for commercial use after the turn of the century, and Liquid Metal Reactors (LWRs) to provide energy production and to convert reactor fission waste to a more manageable waste product. Each technology contributes to the energy solution. Light Water Reactors For The 1990s And Beyond--The U. S. Program The economic and national security of the United States requires a diversified energy supply base built primarily upon adequate, domestic resources that are relatively free from international pressures. Nuclear energy is a vital component of this supply and is essential to meet current and future national energy demands. It is a safe, economically continues to contribute to national energy stability, and strength. The Light Water Reactor (LWR) has been a major and successful contributor to the electrical generating needs of many nations throughout the world. It is being counted upon in the United States as a key to revitalizing nuclear energy option in the 1990s. In recent years, DOE joined with the industry to ensure the availability and future viability of the LWR option. This national program has the participation of the Nation's utility industry, the Electric Power Research Institute (EPRI), and several of the major reactor manufacturers and architect-engineers. Separate but coordinated parts of this program are managed by EPRI and DOE.

A light water excess heat reaction suggests that cold fusion may be alkali-hydrogen fusion

International Nuclear Information System (INIS)

Bush, R.T.

1992-01-01

This paper reports that Mills and Kneizys presented data in support of a light water excess heat reaction obtained with an electrolytic cell highly reminiscent of the Fleischmann-Pons cold fusion cell. The claim of Mills and Kneizys that their excess heat reaction can be explained on the basis of a novel chemistry, which supposedly also explains cold fusion, is rejected in favor of their reaction being, instead, a light water cold fusion reaction. It is the first known light water cold fusion reaction to exhibit excess heat, it may serve as a prototype to expand our understanding of cold fusion. From this new reactions are deduced, including those common to past cold fusion studies. This broader pattern of nuclear reactions is typically seen to involve a fusion of the nuclides of the alkali atoms with the simplest of the alkali-type nuclides, namely, protons, deuterons, and tritons. Thus, the term alkali-hydrogen fusion seems appropriate for this new type of reaction with three subclasses: alkali-hydrogen fusion, alkali-deuterium fusion, and alkali-tritium fusion. A new three-dimensional transmission resonance model (TRM) is sketched. Finally, preliminary experimental evidence in support of the hypothesis of a light water nuclear reaction and alkali-hydrogen fusion is reported. Evidence is presented that appears to strongly implicate the transmission resonance phenomenon of the new TRM

Proceedings of GLOBAL 2007 conference on advanced nuclear fuel cycles and systems

International Nuclear Information System (INIS)

2007-01-01

In keeping with the 12-year history of this conference, GLOBAL 2007 focuses on future nuclear energy systems and fuel cycles. With the increasing public acceptance and political endorsement of nuclear energy, it is a pivotal time for nuclear energy research. Significant advances have been made in development of advanced nuclear fuels and materials, reactor designs, partitioning, transmutation and reprocessing technologies, and waste management strategies. In concert with the technological advances, it is more important than ever to develop sensible nuclear proliferation policies, to promote sustainability, and to continue to increase international collaboration. To further these aims, GLOBAL 2007 highlights recent developments in the following areas: advanced integrated fuel cycle concepts, spent nuclear fuel reprocessing, advanced reprocessing technology, advanced fuels and materials, advanced waste management technology, novel concepts for waste disposal and repository development, advanced reactors, partitioning and transmutation, developments in nuclear non-proliferation technology, policy, and implementation, sustainability and expanded global utilization of nuclear energy, and international collaboration on nuclear energy

Passive Safety Systems in Advanced Water Cooled Reactors (AWCRS). Case Studies. A Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

International Nuclear Information System (INIS)

2013-09-01

This report presents the results from the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) collaborative project (CP) on Advanced Water Cooled Reactor Case Studies in Support of Passive Safety Systems (AWCR), undertaken under the INPRO Programme Area C. INPRO was launched in 2000 - on the basis of a resolution of the IAEA General Conference (GC(44)/RES/21) - to ensure that nuclear energy is available in the 21st century in a sustainable manner, and it seeks to bring together all interested Member States to consider actions to achieve innovation. An important objective of nuclear energy system assessments is to identify 'gaps' in the various technologies and corresponding research and development (R and D) needs. This programme area fosters collaboration among INPRO Member States on selected innovative nuclear technologies to bridge technology gaps. Public concern about nuclear reactor safety has increased after the Fukushima Daiichi nuclear power plant accident caused by the loss of power to pump water for removing residual heat in the core. As a consequence, there has been an increasing interest in designing safety systems for new and advanced reactors that are passive in nature. Compared to active systems, passive safety features do not require operator intervention, active controls, or an external energy source. Passive systems rely only on physical phenomena such as natural circulation, thermal convection, gravity and self-pressurization. Passive safety features, therefore, are increasingly recognized as an essential component of the next-generation advanced reactors. A high level of safety and improved competitiveness are common goals for designing advanced nuclear power plants. Many of these systems incorporate several passive design concepts aimed at improving safety and reliability. The advantages of passive safety systems include simplicity, and avoidance of human intervention, external power or signals. For these reasons, most

Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

International Nuclear Information System (INIS)

Rebak, Raul B.

2014-01-01

The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

2014-09-30

The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

Advances in nuclear science and technology

CERN Document Server

Henley, Ernest J

1962-01-01

Advances in Nuclear Science and Technology, Volume 1 provides an authoritative, complete, coherent, and critical review of the nuclear industry. This book covers a variety of topics, including nuclear power stations, graft polymerization, diffusion in uranium alloys, and conventional power plants.Organized into seven chapters, this volume begins with an overview of the three stages of the operation of a power plant, either nuclear or conventionally fueled. This text then examines the major problems that face the successful development of commercial nuclear power plants. Other chapters consider

New generation of nuclear reactors

International Nuclear Information System (INIS)

Chwaszczewski, S.

2000-01-01

The development trends of the construction of nuclear reactors has been performed on the background of worldwide electricity demand for now and predicted for future. The social acceptance, political and economical circumstances has been also taken into account. Seems to Electric Power Research Institute (US) and other national authorities the advanced light water reactors have the best features and chances for further development and commercial applications in future

Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

Energy Technology Data Exchange (ETDEWEB)

1993-05-13

Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

On the safety performance of the advanced nuclear energy systems

International Nuclear Information System (INIS)

Li Shounan

1999-01-01

Some features on the safety performances of the Advanced Nuclear Energy Systems are discussed. The advantages and some peculiar problems on the safety of Advanced Nuclear Energy Systems with subcritical nuclear reactor driven by external neutron sources are also pointed out in comparison with conventional nuclear reactors

Public concerns and alternative nuclear power systems

International Nuclear Information System (INIS)

Mayo, L.H.

1980-02-01

The basic task undertaken in this study was to assess the relative public acceptability of three general types of nuclear power systems as alternatives to the existing Light Water Reactor (LWR) system. Concerns registered toward nuclear power constituted the basic data for this assessment. The primary measure adopted for determining the significance of concerns was the degree of difficulty posed by the concern to the nuclear power decisional structure in the establishment and maintenance of norms to control risks or to advance intended energy objectives. Alleviations or exacerbations of concern resulting from particular attributes of alternative systems were measured from an LWR baseline

Design of an organic simplified nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Shirvan, Koroush [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States); Forrest, Eric [Primary Standards Laboratory, Sandia National Laboratories, Albuquerque (United States)

2016-08-15

Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

Design of an Organic Simplified Nuclear Reactor

Directory of Open Access Journals (Sweden)

Koroush Shirvan

2016-08-01

Full Text Available Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

[The Chinese nuclear test and 'atoms for peace' as a measure for preventing nuclear armament of Japan: the nuclear non-proliferation policy of the United States and the introduction of light water reactors into Japan, 1964-1968].

Science.gov (United States)

Yamazaki, Masakatsu

2014-07-01

Japan and the United States signed in 1968 a new atomic energy agreement through which US light-water nuclear reactors, including those of the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company, were to be introduced into Japan. This paper studies the history of negotiations for the 1968 agreement using documents declassified in the 1990s in the US and Japan. After the success of the Chinese nuclear test in October 1964, the United States became seriously concerned about nuclear armament of other countries in Asia including Japan. Expecting that Japan would not have its own nuclear weapons, the US offered to help the country to demonstrate its superiority in some fields of science including peaceful nuclear energy to counter the psychological effect of the Chinese nuclear armament. Driven by his own political agenda, the newly appointed Prime Minister Eisaku Sato responded to the US expectation favorably. When he met in January 1965 with President Johnson, Sato made it clear that Japan would not pursue nuclear weapons. Although the US continued its support after this visit, it nevertheless gave priority to the control of nuclear technology in Japan through the bilateral peaceful nuclear agreement. This paper argues that the 1968 agreement implicitly meant a strategic measure to prevent Japan from going nuclear and also a tactic to persuade Japan to join the Nuclear Non -Proliferation Treaty.

Light water detritiation

Energy Technology Data Exchange (ETDEWEB)

Fedorchenko, O.A.; Aleksee, I.A.; Bondarenko, S.D.; Vasyanina, T.V. [B.P. Konstantinov Petersburg Nuclear Physics Institute of National Research Centre ' Kurchatov Institute' , Gatchina (Russian Federation)

2015-03-15

Hundreds of thousands of tons of tritiated light water have been accumulating from the enterprises of nuclear fuel cycles around the world. The Dual-Temperature Water-Hydrogen (DTWH) process looks like the only practical alternative to Combined Electrolysis and Catalytic Exchange (CECE). In DTWH power-consuming lower reflux device (electrolytic cell) is replaced by a so-called 'hot tower' (LPCE column operating at conditions which ensure relatively small value of elementary separation factor α(hot)). In the upper, cold tower, the tritium transfers from hydrogen to water while in the lower, hot tower - in the opposite direction - from water to hydrogen. The DTWH process is much more complicated compared to CECE; it must be thoroughly computed and strictly controlled by an automatic control system. The use of a simulation code for DTWH is absolutely important. The simulation code EVIO-5 deals with 3 flows inside a column (hydrogen gas, water vapour and liquid water) and 2 simultaneous isotope exchange sub-processes (counter-current phase exchange and co-current catalytic exchange). EVIO-5 takes into account the strong dependence of process performance on given conditions (temperature and pressure). It calculates steady-state isotope concentration profiles considering a full set of reversible exchange reactions between different isotope modifications of water and hydrogen (12 molecular species). So the code can be used for simulation of LPCE column operation for detritiation of hydrogen and water feed, which contains H and D not only at low concentrations but above 10 at.% also. EVIO-5 code is used to model a Tritium Removal Facility with a throughput capacity of about 400 m{sup 3}/day. Simulation results show that a huge amount of wet-proofed catalyst is required (about 6000 m{sup 3}), mainly (90%) in the first stage. One reason for these large expenses (apart from a big scale of the problem itself) is the relatively high tritium separation factor in the

Light water detritiation

International Nuclear Information System (INIS)

Fedorchenko, O.A.; Aleksee, I.A.; Bondarenko, S.D.; Vasyanina, T.V.

2015-01-01

Hundreds of thousands of tons of tritiated light water have been accumulating from the enterprises of nuclear fuel cycles around the world. The Dual-Temperature Water-Hydrogen (DTWH) process looks like the only practical alternative to Combined Electrolysis and Catalytic Exchange (CECE). In DTWH power-consuming lower reflux device (electrolytic cell) is replaced by a so-called 'hot tower' (LPCE column operating at conditions which ensure relatively small value of elementary separation factor α(hot)). In the upper, cold tower, the tritium transfers from hydrogen to water while in the lower, hot tower - in the opposite direction - from water to hydrogen. The DTWH process is much more complicated compared to CECE; it must be thoroughly computed and strictly controlled by an automatic control system. The use of a simulation code for DTWH is absolutely important. The simulation code EVIO-5 deals with 3 flows inside a column (hydrogen gas, water vapour and liquid water) and 2 simultaneous isotope exchange sub-processes (counter-current phase exchange and co-current catalytic exchange). EVIO-5 takes into account the strong dependence of process performance on given conditions (temperature and pressure). It calculates steady-state isotope concentration profiles considering a full set of reversible exchange reactions between different isotope modifications of water and hydrogen (12 molecular species). So the code can be used for simulation of LPCE column operation for detritiation of hydrogen and water feed, which contains H and D not only at low concentrations but above 10 at.% also. EVIO-5 code is used to model a Tritium Removal Facility with a throughput capacity of about 400 m 3 /day. Simulation results show that a huge amount of wet-proofed catalyst is required (about 6000 m 3 ), mainly (90%) in the first stage. One reason for these large expenses (apart from a big scale of the problem itself) is the relatively high tritium separation factor in the hot tower

Long term review of research on light water reactor types

International Nuclear Information System (INIS)

Sumiya, Yutaka

1982-01-01

In Japan, 24 nuclear power plants of 17.18 million kWe capacity are in operation, and their rate of operation has shown the good result of more than 60% since 1980. One of the research on the development of light water reactors is the electric power common research, which was started in 1976, and 272 researches were carried out till 1982. It contributed to the counter-measures to stress corrosion cracking, thermal fatigue and the thinning of steam generator tubes, to the reduction of crud generation and the remote control and automation of inspection and maintenance, and to the verification of safety. The important items for the future are the cost down of nuclear power plant construction, the development of robots for nuclear power plants, the improvement of the ability to follow load variation, and the development of light water reactors of new types. It is necessary to diversify the types of reactors to avoid the effect of a serious trouble which may occur in one type of reactors. Tokyo Electric Power Co., Inc., thinks that the Japanese type PWRs having the technical features of KWU type PWRs are desirable for the future development. The compatibility with the condition of installation permission in Japan, the required design change and the economy of the standard design PWRs of KWU (1.3 million kW) have been studied since October, 1981, by KWU and three Japanese manufacturers. (Kako, I.)

Advanced fuels with reduced actinide generation. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1996-11-01

Nuclear energy can play an important future role in supplying the world population with energy. However, this form of energy will be successful only under certain conditions: it must meet very strict safety requirements, it must be economically competitive, and it must be acceptable to the public. Nuclear power produces radioactive wastes and in several countries the public raises concern about safety. Much development work on advanced nuclear power systems is going on in several countries, with participation of both governmental and private industries to meet these conditions. In the framework of this IAEA activity the Technical Committee Meeting on Advanced Fuels with Reduced Actinide Generation was organized. The aim of the meeting was to highlight current research activities and to identify new research areas and fields of possible co-operation. The scope of the meeting included advanced fuels for all types of nuclear reactors: light water reactors, heavy water reactors, high temperature reactors, fast reactors, molten salt reactors and for accelerator driven systems. Other topics covered a wide range of investigations made, or to be made in the Member States. Refs, figs, tabs

Trends in the design of advanced nuclear reactors

International Nuclear Information System (INIS)

Poong-Eil Juhn; Kupitz, Juergen

1996-01-01

Nuclear energy is an essentially unlimited energy source with the potential to provide energy in the form of electricity, district heat and process heat environmentally acceptable conditions. However, this potential will be realized only if nuclear power plants can meet the challenges of national safety requirements, economic competitiveness and public acceptance. Worldwide, a tremendous amount of experience has been accumulated during the development, licensing, construction and operation of nuclear power plants. This experience forms a sound basis for further improvements. Nuclear programmes in the IAEA Member States are addressing the development of advanced reactors, which are intended to have better economics, higher reliability and improved safety. The IAEA, as a global international governmental organization dealing with nuclear power, promotes international information exchange and international co-operation between all countries with their own advanced power programmes and offers assistance to countries with an interest in exploratory or research programmes. The paper gives an overview of global trends in the design of advanced nuclear reactors for electricity generation and heat production along with the role of IAEA. (author)

Advances in commercial heavy water reactor power stations

International Nuclear Information System (INIS)

Brooks, G.L.

1987-01-01

Generating stations employing heavy water reactors have now firmly established an enviable record for reliable, economic electricity generation. Their designers recognize, however, that further improvements are both possible and necessary to ensure that this reactor type remains attractively competitive with alternative nuclear power systems and with fossil-fuelled generation plants. This paper outlines planned development thrusts in a number of important areas, viz., capital cost reduction, advanced fuel cycles, safety, capacity factor, life extension, load following, operator aida, and personnel radiation exposure. (author)

Uranium requirements for advanced fuel cycles in expanding nuclear power systems

International Nuclear Information System (INIS)

Banerjee, S.; Tamm, H.

1978-01-01

When considering advanced fuel cycle strategies in rapidly expanding nuclear power systems, equilibrium analyses do not apply. A computer simulation that accounts for system delay times and fissile inventories has been used to study the effects of different fuel cycles and different power growth rates on uranium consumption. The results show that for a given expansion rate of installed capacity, the main factors that affect resource requirements are the fissile inventory needed to introduce the advanced fuel cycle and the conversion (or breeding) ratio. In rapidly expanding systems, the effect of fissile inventory dominates, whereas in slowly expanding systems, conversion or breeding ratio dominates. Heavy-water-moderated and -cooled reactors, with their high conversion ratios, appear to be adaptable vehicles for accommodating fuel cycles covering a wide range of initial fissile inventories. They are therefore particularly suitable for conserving uranium over a wide range of nuclear power system expansion rates

Press kit. EPR (European pressurized water reactor). The advanced nuclear reactor

International Nuclear Information System (INIS)

2004-10-01

Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21 century, which puts the emphasis on sustainable development. In this framework, this document presents the advantages of the EPR (European Pressurized water Reactor). The EPR is the only third generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. (A.L.B.)

Water chemistry in nuclear power stations with high-temperature reactors with particular reference to the AVR

International Nuclear Information System (INIS)

Nieder, R.; Resch, G.

1976-01-01

The water-steam cycle of a nuclear power plant with a helium-cooled high-temperature reactor differs in design data significantly and extensively from the corresponding cycles of light-water-cooled nuclear reactors and resembles to a great extent the water-steamcycle of a modern conventional power plant. The radioactive constituents of the water-steam cycle can be satisfactorily removed apart from Tritium by means of a pre-coat filter with powder-resisn, as comprehensive experiments have demonstrated. (orig.) [de

Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report

International Nuclear Information System (INIS)

Johansen, R.

2011-01-01

Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

Advanced Light Source Activity Report 2002

Energy Technology Data Exchange (ETDEWEB)

Duque, Theresa; Greiner, Annette; Moxon, Elizabeth; Robinson, Arthur; Tamura, Lori (Editors)

2003-06-12

This annual report of the Advanced Light Source details science highlights and facility improvements during the year. It also offers information on events sponsored by the facility, technical specifications, and staff and publication information.

Advanced Light Source Activity Report 2002

International Nuclear Information System (INIS)

Duque, Theresa; Greiner, Annette; Moxon, Elizabeth; Robinson, Arthur; Tamura, Lori

2003-01-01

This annual report of the Advanced Light Source details science highlights and facility improvements during the year. It also offers information on events sponsored by the facility, technical specifications, and staff and publication information

Electrochemistry in light water reactors reference electrodes, measurement, corrosion and tribocorrosion issues

CERN Document Server

Bosch, R -W; Celis, Jean-Pierre

2007-01-01

There has long been a need for effective methods of measuring corrosion within light water nuclear reactors. This important volume discusses key issues surrounding the development of high temperature reference electrodes and other electrochemical techniques. The book is divided into three parts with part one reviewing the latest developments in the use of reference electrode technology in both pressurised water and boiling water reactors. Parts two and three cover different types of corrosion and tribocorrosion and ways they can be measured using such techniques as electrochemical impedance spectroscopy. Topics covered across the book include in-pile testing, modelling techniques and the tribocorrosion behaviour of stainless steel under reactor conditions. Electrochemistry in light water reactors is a valuable reference for all those concerned with corrosion problems in this key technology for the power industry. Discusses key issues surrounding the development of high temperature reference eletrodes A valuab...

Laser light scattering instrument advanced technology development

Science.gov (United States)

Wallace, J. F.

1993-01-01

The objective of this advanced technology development (ATD) project has been to provide sturdy, miniaturized laser light scattering (LLS) instrumentation for use in microgravity experiments. To do this, we assessed user requirements, explored the capabilities of existing and prospective laser light scattering hardware, and both coordinated and participated in the hardware and software advances needed for a flight hardware instrument. We have successfully breadboarded and evaluated an engineering version of a single-angle glove-box instrument which uses solid state detectors and lasers, along with fiber optics, for beam delivery and detection. Additionally, we have provided the specifications and written verification procedures necessary for procuring a miniature multi-angle LLS instrument which will be used by the flight hardware project which resulted from this work and from this project's interaction with the laser light scattering community.

Aerosol behavior and light water reactor source terms

International Nuclear Information System (INIS)

Abbey, F.; Schikarski, W.O.

1988-01-01

The major developments in nuclear aerosol modeling following the accident to pressurized water reactor Unit 2 at Three Mile Island are briefly reviewed and the state of the art summarized. The importance and implications of these developments for severe accident source terms for light water reactors are then discussed in general terms. The treatment is not aimed at identifying specific source term values but is intended rather to illustrate trends, to assess the adequacy of the understanding of major aspects of aerosol behavior for source term prediction, and demonstrate in qualitative terms the effect of various aspects of reactor design. Areas where improved understanding of aerosol behavior might lead to further reductions in current source terms predictions are also considered

Pilot program to identify valve failures which impact the safety and operation of light water nuclear power plants

International Nuclear Information System (INIS)

Tsacoyeanes, J.C.; Raju, P.P.

1980-04-01

The pilot program described has been initiated under the Department of Energy Light Water Reactor Safety Research and Development Program and has the following specific objectives: to identify the principal types and causes of failures in valves, valve operators and their controls and associated hardware, which lead to, or could lead to plant trip; and to suggest possible remedies for the prevention of these failures and recommend future research and development programs which could lead to minimizing these valve failures or mitigating their effect on plant operation. The data surveyed cover incidents reported over the six-year period, beginning 1973 through the end of 1978. Three sources of information on valve failures have been consulted: failure data centers, participating organizations in the nuclear power industry, and technical documents

Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

International Nuclear Information System (INIS)

1993-01-01

This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval

Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-09-15

This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

Trends in light water reactor dosimetry programs

International Nuclear Information System (INIS)

Rahn, F.J.; Serpan, C.Z.; Fabry, A.; McElroy, W.N.; Grundl, J.A.; Debrue, J.

1977-01-01

Dosimetry programs and techniques play an essential role in the continued assurance of the safety and reliability of components of light water reactors. Primary concern focuses on the neutron irradiation embrittlement of reactor pressure vessels and methods by which the integrity of a pressure vessel can be predicted and monitored throughout its service life. Research in these areas requires a closely coordinated program which integrates the elements of the calculational and material sciences, the development of advanced dosimetric techniques and the use of benchmarks and validation of these methods. The paper reviews the status of the various international efforts in the dosimetry area

Effects of Water Radiolysis in Water Cooled Reactors - Nuclear Energy Research Initiative (NERI) Program. Technical Progress Report

International Nuclear Information System (INIS)

Pimblott, S.M.

2000-01-01

OAK B188 Quarterly Progress Report on NERI Proposal No.99-0010 for the Development of an Experiment and Calculation Based Model to Describe the Effects of Radiation on Non-standard Aqueous Systems Like Those Encountered in the Advanced Light Water Reactor

The economics of the fuel cycle (light water reactors)

International Nuclear Information System (INIS)

Lepine, J.

1979-01-01

The economical characteristics of the fuel cycle (of light water reactors) as well as the definition and calculation method for the average updated cost of the kWh are recalled. The evolution followed by the unit prices of the different operations of the cycle, their total cost and the part taken by this cost in the overall cost of nuclear kWh are described. The effects on the cost of fuel of certain hypotheses, operating requirements and additional cost factors are considered [fr

Advanced spent fuel processing technologies for the United States GNEP programme

International Nuclear Information System (INIS)

Laidler, J.J.

2007-01-01

Spent fuel processing technologies for future advanced nuclear fuel cycles are being developed under the scope of the Global Nuclear Energy Partnership (GNEP). This effort seeks to make available for future deployment a fissile material recycling system that does not involve the separation of pure plutonium from spent fuel. In the nuclear system proposed by the United States under the GNEP initiative, light water reactor spent fuel is treated by means of a solvent extraction process that involves a group extraction of transuranic elements. The recovered transuranics are recycled as fuel material for advanced burner reactors, which can lead in the long term to fast reactors with conversion ratios greater than unity, helping to assure the sustainability of nuclear power systems. Both aqueous and pyrochemical methods are being considered for fast reactor spent fuel processing in the current US development programme. (author)

Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

International Nuclear Information System (INIS)

Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

1980-06-01

The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and 233 U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles

Microencapsulated fuel technology for commercial light water and advanced reactor application

International Nuclear Information System (INIS)

Terrani, Kurt A.; Snead, Lance L.; Gehin, Jess C.

2012-01-01

The potential application of microencapsulated fuels to light water reactors (LWRs) has been explored. The specific fuel manifestation being put forward is for coated fuel particles embedded in silicon carbide or zirconium metal matrices. Detailed descriptions of these concepts are presented, along with a review of attributes, potential benefits, and issues with respect to their application in LWR environments, specifically from the standpoints of materials, neutronics, operations, and economics. Preliminary experiment and modeling results imply that with marginal redesign, significant gains in operational reliability and accident response margins could be potentially achieved by replacing conventional oxide-type LWR fuel with microencapsulated fuel forms.

Report from the Light Water Reactor Sustainability Workshop on On-Line Monitoring Technologies

International Nuclear Information System (INIS)

Baldwin, Thomas; Tawfik, Magdy; Bond, Leonard

2010-01-01

In support of expanding the use of nuclear power, interest is growing in methods of determining the feasibility of longer term operation for the U.S. fleet of nuclear power plants, particularly operation beyond 60 years. To help establish the scientific and technical basis for such longer term operation, the DOE-NE has established a research and development (R and D) objective. This objective seeks to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors. The Light Water Reactor Sustainability (LWRS) Program, which addresses the needs of this objective, is being developed in collaboration with industry R and D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of nuclear power plants. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. In moving to identify priorities and plan activities, the Light Water Reactor Sustainability Workshop on On-Line Monitoring (OLM) Technologies was held June 10-12, 2010, in Seattle, Washington. The workshop was run to enable industry stakeholders and researchers to identify the nuclear industry needs in the areas of future OLM technologies and corresponding technology gaps and research capabilities. It also sought to identify approaches for collaboration that would be able to bridge or fill the technology gaps. This report is the meeting proceedings, documenting the presentations and discussions of the workshop and is intended to serve as a basis for a plan which is under development that will enable the I and C research pathway to achieve its goals. Benefits to the nuclear industry accruing from On Line Monitoring Technology cannot be ignored. Information gathered thus far has contributed significantly to the Department of Energy's Light Water Reactor Sustainability Program. DOE

Report from the Light Water Reactor Sustainability Workshop on On-Line Monitoring Technologies

Energy Technology Data Exchange (ETDEWEB)

Thomas Baldwin; Magdy Tawfik; Leonard Bond

2010-06-01

In support of expanding the use of nuclear power, interest is growing in methods of determining the feasibility of longer term operation for the U.S. fleet of nuclear power plants, particularly operation beyond 60 years. To help establish the scientific and technical basis for such longer term operation, the DOE-NE has established a research and development (R&D) objective. This objective seeks to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors. The Light Water Reactor Sustainability (LWRS) Program, which addresses the needs of this objective, is being developed in collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of nuclear power plants. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. In moving to identify priorities and plan activities, the Light Water Reactor Sustainability Workshop on On-Line Monitoring (OLM) Technologies was held June 10–12, 2010, in Seattle, Washington. The workshop was run to enable industry stakeholders and researchers to identify the nuclear industry needs in the areas of future OLM technologies and corresponding technology gaps and research capabilities. It also sought to identify approaches for collaboration that would be able to bridge or fill the technology gaps. This report is the meeting proceedings, documenting the presentations and discussions of the workshop and is intended to serve as a basis for a plan which is under development that will enable the I&C research pathway to achieve its goals. Benefits to the nuclear industry accruing from On Line Monitoring Technology cannot be ignored. Information gathered thus far has contributed significantly to the Department of Energy’s Light Water Reactor Sustainability Program. DOE has

Nuclear astrophysics of light nuclei

DEFF Research Database (Denmark)

Fynbo, Hans Otto Uldall

2013-01-01

A review of nuclear astrophysics of light nuclei using radioactive beams or techniques developed for radioactive beams is given. We discuss Big Bang nucleosynthesis, with special focus on the lithium problem, aspects of neutrino-physics, helium-burning and finally selected examples of studies...

The advanced light source: America's brightest light for science and industry

International Nuclear Information System (INIS)

Cross, J.; Lawler, G.

1994-03-01

America's brightest light comes from the Advanced Light Source (ALS), a national facility for scientific research, product development, and manufacturing. Completed in 1993, the ALS produces light in the ultraviolet and x-ray regions of the spectrum. Its extreme brightness provides opportunities for scientific and technical progress not possible anywhere else. Technology is poised on the brink of a major revolution - one in which vital machine components and industrial processes will be drastically miniaturized. Industrialized nations are vying for leadership in this revolution - and the huge economic rewards the leaders will reap

Recent advances in nuclear cardiology

DEFF Research Database (Denmark)

Gutte, H.; Petersen, C. Leth; Kjaer, A.

2008-01-01

Nuclear cardiology is an essential part of functional, non-invasive, cardiac imaging. Significant advances have been made in nuclear cardiology since planar (201)thallium ((201)TI) scintigraphy was introduced for the evaluation of left ventricular (LV) perfusion nearly 40 years ago. The use...... of nuclear cardiology has been steadily increasing over the last 20 years with important steps being the introduction of (99m)technetium- ((99m)Tc)-labelled perfusion radiotracers, the change from only planar to now much more single photon emission computed tomography (SPECT) and positron emission tomography...... (PET), electrocardiogram gating of nuclear perfusion imaging, and finally introducing nuclear hybrid imaging using either SPECT or PET together with either computed tomography or magnetic resonance imaging. The indications have extended from nearly only coronary artery diseases to several non...

Advanced Fuel Pellet Materials and Fuel Rod Design for Water Cooled Reactors. Proceedings of a Technical Committee Meeting

International Nuclear Information System (INIS)

2010-10-01

The economics of current nuclear power plants have improved through increased fuel burnup and longer fuel cycles, i.e. increasing the effective time that fuel remains in the reactor core and the amount of energy it generates. Efficient consumption of fissile material in the fuel element before it is discharged from the reactor means that less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. Better utilization of fissile nuclear materials, as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all, on fuel and cladding materials. It entails increased attention to measures ensuring desired in-pile fuel performance parameters that require adequate improvements in fuel material properties and fuel rod designs. These are the main reasons that motivated the IAEA Technical Working Group on Fuel Performance and Technology (TWG-FPT) to recommend the organization of a Technical Committee Meeting on Advanced Fuel Pellet Materials and Fuel Rod Designs for Power Reactors. The proposal was supported by the IAEA TWGs on Advanced Technologies for Light and Heavy Water-Cooled Reactors (TWG-LWR and TWG-HWR), and the meeting was held at the invitation of the Government of Switzerland at the Paul Scherrer Institute in Villigen, from 23 to 26 November 2009. This was the third IAEA meeting on these subjects (the first was held in 1996 in Tokyo, Japan, and the second in 2003 in Brussels, Belgium), which reflects the continuous interest in the above issues among Member States. The purpose of the meeting was to review the current status in the development of fuel pellet materials and to explore recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting was attended by 45 specialists representing fuel vendors, nuclear utilities, research and development

US nuclear power programs

International Nuclear Information System (INIS)

McGolf, D.J.

1994-01-01

In the United States, coal provided 56 percent of the electricity generated in 1992. Nuclear energy was the next largest contributor, supplying 22 percent. Natural gas provided 9 percent, while hydro-electric and renewables together supplied another 9 percent. Currently, the 109 nuclear power plants in the U.S. have an overall generating capacity of 99,000 MWe. To improve efficiency, safety, and performance, the lessons of 30 years of experience with nuclear powerplants are being incorporated into design criteria for the next generation of U.S. plants. The new Advanced Light Water Reactor plants will feature simpler designs, which will enable more cost-effective construction and maintenance. To enhance safety, design margins are being increased, and human factors are being considered and incorporated into the designs

US nuclear power programs

Energy Technology Data Exchange (ETDEWEB)

McGolf, D J

1994-12-31

In the United States, coal provided 56 percent of the electricity generated in 1992. Nuclear energy was the next largest contributor, supplying 22 percent. Natural gas provided 9 percent, while hydro-electric and renewables together supplied another 9 percent. Currently, the 109 nuclear power plants in the U.S. have an overall generating capacity of 99,000 MWe. To improve efficiency, safety, and performance, the lessons of 30 years of experience with nuclear powerplants are being incorporated into design criteria for the next generation of U.S. plants. The new Advanced Light Water Reactor plants will feature simpler designs, which will enable more cost-effective construction and maintenance. To enhance safety, design margins are being increased, and human factors are being considered and incorporated into the designs.

Advanced Light Source Activity Report 2000

International Nuclear Information System (INIS)

Greiner, A.; Moxon, L.; Robinson, A.; Tamura, L.

2001-01-01

This is an annual report, detailing activities at the Advanced Light Source for the year 2000. It includes highlights of scientific research by users of the facility as well as information about the development of the facility itself

Advanced Light Source Activity Report 2000

Energy Technology Data Exchange (ETDEWEB)

Greiner, A.; Moxon, L.; Robinson, A.; Tamura, L.

2001-04-01

This is an annual report, detailing activities at the Advanced Light Source for the year 2000. It includes highlights of scientific research by users of the facility as well as information about the development of the facility itself.

Technology programs in support of advanced light water reactors

International Nuclear Information System (INIS)

1990-03-01

Design for Constructability's overall purpose was to identify and address changes in the nuclear industry to restore nuclear energy as an attractive option for new generating capacity. The program stove to meet the following goals related to the future construction of nuclear power plants: reduced costs; assurance of improved quality; and shortened construction schedules. This is Volume 1 of three volumes

Analysis of thermal fatigue events in light water reactors

Energy Technology Data Exchange (ETDEWEB)

Okuda, Yasunori [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

2000-09-01

Thermal fatigue events, which may cause shutdown of nuclear power stations by wall-through-crack of pipes of RCRB (Reactor Coolant Pressure Boundary), are reported by licensees in foreign countries as well as in Japan. In this paper, thermal fatigue events reported in anomalies reports of light water reactors inside and outside of Japan are investigated. As a result, it is clarified that the thermal fatigue events can be classified in seven patterns by their characteristics, and the trend of the occurrence of the events in PWRs (Pressurized Water Reactors) has stronger co-relation to operation hours than that in BWRs (Boiling Water Reactors). Also, it is concluded that precise identification of locations where thermal fatigue occurs and its monitoring are important to prevent the thermal fatigue events by aging or miss modification. (author)

Advances in nuclear science and technology

CERN Document Server

Henley, Ernest J

1976-01-01

Advances in Nuclear Science and Technology, Volume 9 provides information pertinent to the fundamental aspects of nuclear science and technology. This book discusses the safe and beneficial development of land-based nuclear power plants.Organized into five chapters, this volume begins with an overview of the possible consequences of a large-scale release of radioactivity from a nuclear reactor in the event of a serious accident. This text then discusses the extension of conventional perturbation techniques to multidimensional systems and to high-order approximations of the Boltzmann equation.

Advances in nuclear science and technology

CERN Document Server

Henley, Ernest J

1972-01-01

Advances in Nuclear Science and Technology, Volume 6 provides information pertinent to the fundamental aspects of nuclear science and technology. This book covers a variety of topics, including nuclear steam generator, oscillations, fast reactor fuel, gas centrifuge, thermal transport system, and fuel cycle.Organized into six chapters, this volume begins with an overview of the high standards of technical safety for Europe's first nuclear-propelled merchant ship. This text then examines the state of knowledge concerning qualitative results on the behavior of the solutions of the nonlinear poin

The advanced controls program at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Knee, H.E.; White, J.D.

1990-01-01

The Oak Ridge National Laboratory (ORNL), under sponsorship of the US Department of Energy (DOE), is conducting research that will lead to advanced, automated control of new liquid-metal-reactor (LMR) nuclear power plants. Although this program of research (entitled the ''Advanced Controls Program'') is focused on LMR technology, it will be capable of providing control design, test, and qualification capability for other advanced reactor designs (e.g., the advanced light water reactor [ALWR] and high temperature gas-cooled reactor [HTGR] designs), while also benefiting existing nuclear plants. The Program will also have applicability to complex, non-nuclear process control environments (e.g., petrochemical, aerospace, etc.). The Advanced Controls Program will support capabilities throughout the entire plant design life cycle, i.e., from the initial interactive first-principle dynamic model development for the process, systems, components, and instruments through advanced control room qualification. The current program involves five principal areas of research activities: (1) demonstrations of advanced control system designs, (2) development of an advanced controls design environment, (3) development of advanced control strategies, (4) research and development (R ampersand D) in human-system integration for advanced control system designs, and (5) testing and validation of advanced control system designs. Discussion of the research in these five areas forms the basis of this paper. Also included is a description of the research directions of the program. 8 refs

Nuclear energy cost data base: A reference data base for nuclear and coal-fired powerplant power generation cost analysis

International Nuclear Information System (INIS)

1988-09-01

A reference data base and standard methodology are needed for performing comparative nuclear and fossil power generation cost analyses for the Department of Energy, Office of Nuclear Energy. This report contains such a methodology together with reference assumptions and data to be used with the methodology. It is intended to provide basic guidelines or a starting point for analyses and to serve as a focal point in establishing parameters and methods to be used in economic comparisons of nuclear systems with alternatives. The data base is applicable for economic comparisons of new base load light-water reactors on a once-through cycle, and high- and low-sulfur coal-fired plants, and oil- and natural gas-fired electric generating plants coming on line around the turn of the century. In addition to current generation light-water reactors and fossil fuel-fired plants, preliminary cost information is also presented on improved and advanced light-water reactors, liquid metal reactor plants and fuel cycle facilities. This report includes an updated data base containing proposed technical and economic assumptions to be used in analyses, discussions of a recommended methodology to be used in calculating power generation costs, a sample calculation for illustrative and benchmark purposes and projected power generation costs for fission and coal-fired alternatives. Effects of the 1986 Tax Reform Act are included. 126 refs., 17 figs., 47 tabs

Prospects for plutonium recycling in light-water reactors in the European Communities

International Nuclear Information System (INIS)

Haijtink, B.

1977-01-01

On the basis of a forecast on nuclear power installation programmes and spent nuclear fuel reprocessing capacities in the European Community, an assessment has been made of the developmnt of the plutonium handling industry up to the year 1990. Data on plutonium recovered in the reprocessing plants, requirements for fuelling the fast-breeder reactors and the remaining stockpile have been evaluated. A possible strategy for recycling the plutonium surplus in light-water reactors has been developed and its effects as regards saving on requirements for natural uranium and enrichment services have been estimated

Advanced Nuclear Magnetic Resonance

OpenAIRE

Alonso, Diego A.

2014-01-01

Transparencias en inglés de la asignatura "Resonancia Magnética Nuclear Avanzada" (Advanced Nuclear Magnetic Resonance) (36643) que se imparte en el Máster de Química Médica como asignatura optativa de 3 créditos ECTS. En esta asignatura se completa el estudio iniciado en la asignatura de quinto curso de la licenciatura en Química "Determinación estructural" (7448) y en la del Grado de Química de tercer curso "Determinación estructural de los compuestos orgánicos" (26030) en lo referente a té...

Modern light-water reactors - EPR and SWR 1000. Present status and possibilities of development and application

International Nuclear Information System (INIS)

Brettschuh, W.; Schneider, D.

2001-01-01

A number of reactor concepts are being advanced worldwide also in the light of long-term perspectives in the power economy. Activities in this respect are concentrated on the proven technologies of light-water and heavy-water reactors and on gas or liquid-metal cooled designs. Also smaller units are being considered for various applications. Framatome ANP focus on further development of the EPR and SWR 1000 light-water reactor lines. These plant concepts, which were developed in cooperation with, and with the support of, power utilities together with a number of European partners, are characterized by an even higher level of safety, especially as far as potential accidental releases are concerned, and by their competitiveness in the electricity generating market compared to other sources of energy. Given the cost structure of these plants, with approx. 60% capital costs, approx. 24% operating costs, and approx. 16% fuel cycle costs, efforts are made in particular to reduce the dominating capital cost share. For the EPR, this is to be achieved by high plant capacities, i.e. lower specific capital outlays per kilowatt installed; for the SWR 1000, mainly by means of an optimized, streamlined plant design. Moreover, installed nuclear generating capacities provide long-term assurance in calculating electricity generating costs, as they are less sensitive to the variable costs of the fuel cycle and of operation than fossil-fired plants. Commercial maturity of the SWR 1000 will be achieved after the completion of basic development in late 1999 and the ongoing detailed design steps in 2001. Also for the EPR, commercial maturity has been achieved with the completion of the basic design phase. (orig.) [de

Passive safety systems and natural circulation in water cooled nuclear power plants

International Nuclear Information System (INIS)

2009-11-01

Nuclear power produces 15% of the world's electricity. Many countries are planning to either introduce nuclear energy or expand their nuclear generating capacity. Design organizations are incorporating both proven means and new approaches for reducing the capital costs of their advanced designs. In the future most new nuclear plants will be of evolutionary design, often pursuing economies of scale. In the longer term, innovative designs could help to promote a new era of nuclear power. Since the mid-1980s it has been recognized that the application of passive safety systems (i.e. those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially improve economics of new nuclear power plant designs. The IAEA Conference on The Safety of Nuclear Power: Strategy for the Future, which was convened in 1991, noted that for new plants 'the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate'. Some new designs also utilize natural circulation as a means to remove core power during normal operation. The use of passive systems can eliminate the costs associated with the installation, maintenance, and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. To support the development of advanced water cooled reactor designs with passive systems, investigations of natural circulation are conducted in several IAEA Member States with advanced reactor development programmes. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, the IAEA