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Sample records for argonne heavy water modified reactor

  1. Advances in heavy water reactors

    International Nuclear Information System (INIS)

    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

  2. Safety systems of heavy water reactors and small power reactors

    International Nuclear Information System (INIS)

    After introductional descriptions of heavy water reactors and natural circulation boiling water reactors the safety philosophy and safety systems like ECCS, residual heat removal, protection systems etc., are described. (RW)

  3. Heavy water cycle in the CANDU reactor

    International Nuclear Information System (INIS)

    Hydrogen atom has two isotopes: deuterium 1H2 and tritium 1H3. The deuterium oxide D2O is called heavy water due to its density of 1105.2 Kg/m3. Another important physical property of the heavy water is the low neutron capture section, suitable to moderate the neutrons into natural uranium fission reactor as CANDU. Due to the fact that into this reactor the fuel is cooled into the pressure tubes surrounded by a moderator, the usage of D2O as primary heat transport (PHT) agent is mandatory. Therefore a large amount of heavy water (approx. 500 tons) is used in a CANDU reactor. Being a costly resource - it represents 20% of the initial plant capital cost, D2O management is required to preserve it. (author)

  4. Behavior of tritium in heavy water reactors

    International Nuclear Information System (INIS)

    In the ATR Fugen power station, the radiation control regarding the tritium in heavy water has been carried out since the heavy water was filled in the system of the reactor in November, 1977. At first, the concentration of tritium in heavy water was about 60 μCi/cc, but in November, 1981, it increased to about 1.3 mCi/cc, and the saturation concentration after 30 years is estimated to become about 17 mCi/cc. In this report, on the transfer of tritium to the work environment and general environment, its barrier, recovery, measurement and the protection against it, the experience in the Fugen power station is described. The heavy water system was constructed as the perfectly closed circuit by welding stainless steel, and a canned heavy water circulating pump has been used. The leak of heavy water in the steady operation is negligible, but attention must be paid to the transfer of tritium to the environment when the system is disassembled for the regular inspection. The measurement of tritium for individual exposure control, environment and released radioactivity, the tritium-removing equipment and protective suits, and the release of tritium to general environment are reported. (Kako, I.)

  5. Good practices in heavy water reactor operation

    International Nuclear Information System (INIS)

    The value and importance of organizations in the nuclear industry engaged in the collection and analysis of operating experience and best practices has been clearly identified in various IAEA publications and exercises. Both facility safety and operational efficiency can benefit from such information sharing. Such sharing also benefits organizations engaged in the development of new nuclear power plants, as it provides information to assist in optimizing designs to deliver improved safety and power generation performance. In cooperation with Atomic Energy of Canada, Ltd, the IAEA organized the workshop on best practices in Heavy Water Reactor Operation in Toronto, Canada from 16 to 19 September 2008, to assist interested Member States in sharing best practices and to provide a forum for the exchange of information among participating nuclear professionals. This workshop was organized under Technical Cooperation Project INT/4/141, on Status and Prospects of Development for and Applications of Innovative Reactor Concepts for Developing Countries. The workshop participants were experts actively engaged in various aspects of heavy water reactor operation. Participants presented information on activities and practices deemed by them to be best practices in a particular area for consideration by the workshop participants. Presentations by the participants covered a broad range of operational practices, including regulatory aspects, the reduction of occupational dose, performance improvements, and reducing operating and maintenance costs. This publication summarizes the material presented at the workshop, and includes session summaries prepared by the chair of each session and papers submitted by the presenters

  6. Piping installation for reactor heavy water system

    International Nuclear Information System (INIS)

    Characteristics and main installation steps for the piping of the reactor heavy water loop system were introduced in this paper. According to the system design, equipment accommodation and spot management, main issues with effect on the quality and schedule of pipeline installation were analyzed. Accordingly, some solutions were put forward, which included: work allocation should be made clear in documents; installation preparative such as design checkup and technology communication should be prepared completely; requirements of system cleaning, test items in every experiment, inspection in work and equipment maintenance should be considered in the system design; perfect documents distribution system and stock plan should be built; technology requirements and quality assurance should be claimed in contracts; quality should be controlled by way of external evidence, inspection in manufactory, exterior quality assurance examination, and test during consignment; series of management procedure should be established in detail. (authors)

  7. Heavy water moderated reactors advances and challenges

    International Nuclear Information System (INIS)

    Nuclear energy is now considered a key contributor to world electricity production, with total installed capacity nearly equal to that of hydraulic power. Nevertheless, many important challenges lie ahead. Paramount among these is gaining public acceptance: this paper makes the basic assumption that public acceptance will improve if, and only if, nuclear power plants are operated safely and economically over an extended period of time. The first task, therefore, is to ensure that these prerequisites to public acceptance are met. Other issues relate to the many aspects of economics associated with nuclear power, include capital cost, operation cost, plant performance and the risk to the owner's investment. Financing is a further challenge to the expansion of nuclear power. While the ability to finance a project is strongly dependent on meeting public acceptance and economic challenges, substantial localisation of design and manufacture is often essential to acceptance by the purchaser. The neutron efficient heavy water moderated CANDU with its unique tube reactor is considered to be particularly well qualified to respond to these market challenges. Enhanced safety can be achieved through simplification of safety systems, design of the moderator and shield water systems to mitigate severe accident events, and the increased use of passive systems. Economics are improved through reduction in both capital and operating costs, achieved through the application of state-of-the-art technologies and economy of scale. Modular features of the design enhance the potential for local manufacture. Advanced fuel cycles offer reduction in both capital costs and fuelling costs. These cycles, including slightly enriched uranium and low grade fuels from reprocessing plants can serve to increase reactor output, reduce fuelling cost and reduce waste production, while extending resource utilisation. 1 ref., 1 tab

  8. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  9. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    International Nuclear Information System (INIS)

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs

  10. Computer code for the analyses of reactivity initiated accident of heavy water moderated and cooled research reactor 'EUREKA-2D'

    International Nuclear Information System (INIS)

    Codes, such as EUREKA and EUREKA-2 have been developed to analyze the reactivity initiated accident for light water reactor. These codes could not be applied directly for the analyses of heavy water moderated and cooled research reactor which are different from light water reactor not only on operation condition but also on reactor kinetic constants. EUREKA-2D which is modified EUREKA-2 is a code for the analyses of reactivity initiated accident of heavy water research reactors. Following items are modified: 1) reactor kinetic constants. 2) thermodynamic properties of coolant. 3) heat transfer equations. The feature of EUREKA-2D and an example of analysis are described in this report. (author)

  11. Advanced light and heavy water reactors for improved fuel utilization

    International Nuclear Information System (INIS)

    On 26-29 November 1984 the Agency convened at its Headquarters in Vienna the Technical Committee and Workshop on Advanced Light and Heavy Water Reactor Technology in order to provide an opportunity to review and discuss the current status and recent development in the lay-out and design of advanced water reactor and to identify areas in which additional research and development are needed. The meeting was attended by 45 participants from 16 nations and 2 international organizations presenting 25 papers. The Conference presentations were divided into sessions devoted to the following topics: Advanced light water reactor programmes (6 papers); Advanced light water design, technology and physics (12 papers); Advanced heavy water reactors (7 papers). A separate abstract was prepared for each of these papers

  12. Conceptual design of a large heavy water reactor for US siting

    Energy Technology Data Exchange (ETDEWEB)

    Shapiro, N L; Jesick, J F

    1979-09-01

    Information is presented concerning fuel management and safety and licensing assessment of the pressurized heavy water reactor; and commercial introduction of the pressurized heavy water reactor in the United States.

  13. Conceptual design of a large heavy water reactor for US siting

    International Nuclear Information System (INIS)

    Information is presented concerning fuel management and safety and licensing assessment of the pressurized heavy water reactor; and commercial introduction of the pressurized heavy water reactor in the United States

  14. Heavy-Water Power Reactors. Proceedings Of A Symposium

    International Nuclear Information System (INIS)

    Proceedings of a Symposium organized by the IAEA and held in Vienna, 11-15 September 1967. The timeliness of the meeting was underlined by the large gathering of over 225 participants from 28 countries and three international organizations. Contents: Experience with heavy-water power and experimental reactors and projects (14 papers); New and advanced power reactor designs and concepts (8 papers); Development programmes and thorium cycle (9 papers); Economics and prospects of heavy-water power reactors (7 papers); Physics and fuel management (8 papers); Fuels (5 papers); Safety, control and engineering (6 papers); Panel discussion. Except for one Russian paper, which is published in English, each paper is in its original language (49 English and 8 French) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  15. Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors

    OpenAIRE

    Peel Ross; Van Den Durpel Luc; Ogden Mark Daniel; Whittle Karl Rhys

    2016-01-01

    This paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom. Heavy water reactors are used extensively in Canada, and are deploying within India and China, whilst the UK is considering the use of heavy water reactors to manage its plutonium inventory of 140 tonnes. The UK heavy water reactor proposal uses a mix...

  16. Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors

    OpenAIRE

    Peel, R.; Van Den Derpel, L.; Whittle, K.; Ogden, M.

    2016-01-01

    This paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom. Heavy water reactors are used extensively in Canada, and are deploying within India and China, whilst the UK is considering the use of heavy water reactors to manage its plutonium inventory of 140 tonnes. The UK heavy water reactor proposal uses a mix...

  17. Heavy-water-moderated pressure-tube reactor safety

    International Nuclear Information System (INIS)

    Several countries have heavy-water-moderated, pressure-tube reactors either in commercial operation or in late prototype stages. The supporting safety research and development includes such areas as the thermohydraulics of circuit depressurization, heat transfer from the fuel, heat rejection to the moderator from dry fuel, fuel and sheath behaviour, and fuel channel integrity. We review the work done in Canada, Great Britain, Italy and Japan, and describe some of the experimental tests underlaying the methods of accident analysis. The reactors have safety systems which, in the event of an accident, are able to shut down the reactor, keep the fuel cooled, and contain any released radioactivity. We summarize the characteristics of these safety systems (shutdown, emergency core cooling, and containment) in the various reactors, and discuss other reactor characteristics which either prevent accidents or reduce their potential demand on the safety systems. (author)

  18. Antineutrino monitoring for the Iranian heavy water reactor

    CERN Document Server

    Christensen, Eric; Jaffke, Patrick; Shea, Thomas

    2014-01-01

    In this note we discuss the potential application of antineutrino monitoring to the Iranian heavy water reactor at Arak, the IR-40, as a non-proliferation measure. We demonstrate that an above ground detector positioned right outside the IR-40 reactor building could meet and in some cases significantly exceed the verification goals identified by IAEA for plutonium production or diversion from declared inventories. In addition to monitoring the reactor during operation, observing antineutrino emissions from long-lived fission products could also allow monitoring the reactor when it is shutdown. Antineutrino monitoring could also be used to distinguish different levels of fuel enrichment. Most importantly, these capabilities would not require a complete reactor operational history and could provide a means to re-establish continuity of knowledge in safeguards conclusions should this become necessary.

  19. Topical papers on heavy water, fuel fabrication and reactors

    International Nuclear Information System (INIS)

    A total of four papers is presented. The first contribution of the Federal Republic of Germany reviews the market situation for reactors and the relations between reactor producers and buyers as reflected in sales agreements. The second West German contribution gives a world-wide survey of fuel element production as well as of fuel and fuel element demand up to the year 2000. The Canadian paper discusses the future prospects of heavy-water production, while the Ecuador contribution deals with small and medium-sized nuclear power plants

  20. Evaluation of Heavy Metal Removal from Wastewater in a Modified Packed Bed Biofilm Reactor

    OpenAIRE

    Shohreh Azizi; Ilunga Kamika; Memory Tekere

    2016-01-01

    For the effective application of a modified packed bed biofilm reactor (PBBR) in wastewater industrial practice, it is essential to distinguish the tolerance of the system for heavy metals removal. The industrial contamination of wastewater from various sources (e.g. Zn, Cu, Cd and Ni) was studied to assess the impacts on a PBBR. This biological system was examined by evaluating the tolerance of different strengths of composite heavy metals at the optimum hydraulic retention time (HRT) of 2 h...

  1. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition

  2. Reliability analysis of Indian pressurized heavy water reactor piping

    International Nuclear Information System (INIS)

    In this paper, a probabilistic analysis of primary heat transport of Indian Pressurized Heavy Water reactor is presented. The probability of failure of the straight pipes with through wall circumferential flaws subjected to bending moment is calculated. The failure criteria considered is net section collapse and R6 method. Probability of failure is obtained with crack growth initiation as the limiting condition. The variability in the crack size and material properties (tensile and fracture) is considered. The probability of failure is calculated at different levels of applied load. Various methods of probability estimation are presented and their equivalence is demonstrated. The probability of failure is obtained using classical Monte Carlo method, Monte Carlo with importance sampling, First Order Reliability Method (FORM), Second Order Reliability Method (SORM) and by numerical integration of the failure integral using Lepage's VEGAS algorithm. The results are utilized for demonstrating that for the leakage size crack, the pipe design has high probability for leak before break. (orig.)

  3. Evaluation of Heavy Metal Removal from Wastewater in a Modified Packed Bed Biofilm Reactor.

    Science.gov (United States)

    Azizi, Shohreh; Kamika, Ilunga; Tekere, Memory

    2016-01-01

    For the effective application of a modified packed bed biofilm reactor (PBBR) in wastewater industrial practice, it is essential to distinguish the tolerance of the system for heavy metals removal. The industrial contamination of wastewater from various sources (e.g. Zn, Cu, Cd and Ni) was studied to assess the impacts on a PBBR. This biological system was examined by evaluating the tolerance of different strengths of composite heavy metals at the optimum hydraulic retention time (HRT) of 2 hours. The heavy metal content of the wastewater outlet stream was then compared to the source material. Different biomass concentrations in the reactor were assessed. The results show that the system can efficiently treat 20 (mg/l) concentrations of combined heavy metals at an optimum HRT condition (2 hours), while above this strength there should be a substantially negative impact on treatment efficiency. Average organic reduction, in terms of the chemical oxygen demand (COD) of the system, is reduced above the tolerance limits for heavy metals as mentioned above. The PBBR biological system, in the presence of high surface area carrier media and a high microbial population to the tune of 10 000 (mg/l), is capable of removing the industrial contamination in wastewater. PMID:27186636

  4. Evaluation of Heavy Metal Removal from Wastewater in a Modified Packed Bed Biofilm Reactor.

    Directory of Open Access Journals (Sweden)

    Shohreh Azizi

    Full Text Available For the effective application of a modified packed bed biofilm reactor (PBBR in wastewater industrial practice, it is essential to distinguish the tolerance of the system for heavy metals removal. The industrial contamination of wastewater from various sources (e.g. Zn, Cu, Cd and Ni was studied to assess the impacts on a PBBR. This biological system was examined by evaluating the tolerance of different strengths of composite heavy metals at the optimum hydraulic retention time (HRT of 2 hours. The heavy metal content of the wastewater outlet stream was then compared to the source material. Different biomass concentrations in the reactor were assessed. The results show that the system can efficiently treat 20 (mg/l concentrations of combined heavy metals at an optimum HRT condition (2 hours, while above this strength there should be a substantially negative impact on treatment efficiency. Average organic reduction, in terms of the chemical oxygen demand (COD of the system, is reduced above the tolerance limits for heavy metals as mentioned above. The PBBR biological system, in the presence of high surface area carrier media and a high microbial population to the tune of 10 000 (mg/l, is capable of removing the industrial contamination in wastewater.

  5. Evaluation of Heavy Metal Removal from Wastewater in a Modified Packed Bed Biofilm Reactor.

    Science.gov (United States)

    Azizi, Shohreh; Kamika, Ilunga; Tekere, Memory

    2016-01-01

    For the effective application of a modified packed bed biofilm reactor (PBBR) in wastewater industrial practice, it is essential to distinguish the tolerance of the system for heavy metals removal. The industrial contamination of wastewater from various sources (e.g. Zn, Cu, Cd and Ni) was studied to assess the impacts on a PBBR. This biological system was examined by evaluating the tolerance of different strengths of composite heavy metals at the optimum hydraulic retention time (HRT) of 2 hours. The heavy metal content of the wastewater outlet stream was then compared to the source material. Different biomass concentrations in the reactor were assessed. The results show that the system can efficiently treat 20 (mg/l) concentrations of combined heavy metals at an optimum HRT condition (2 hours), while above this strength there should be a substantially negative impact on treatment efficiency. Average organic reduction, in terms of the chemical oxygen demand (COD) of the system, is reduced above the tolerance limits for heavy metals as mentioned above. The PBBR biological system, in the presence of high surface area carrier media and a high microbial population to the tune of 10 000 (mg/l), is capable of removing the industrial contamination in wastewater.

  6. Containment for Heavy-Water Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    The safety principles applicable to heavy-water, gas-cooled reactors are outlined, with a view to establishing containment specifications adapted to the sites available in Switzerland for the construction of nuclear plants. These specifications are derived from dose rates considered acceptable, in the event of a serious reactor accident, for persons living near the plant, and are based on-meteorological and demographic conditions representative of the majority of the country's sites. The authors consider various designs for the containment shell, taking into account the conditions which would exist in the shell after the maximum credible accident. The following types of shell are studied: pre-stressed concrete; pre-stressed concrete with steel dome; pre-stressed concrete with inner, leakproof steel lining; steel with concrete side shield to protect against radiation; double shell. The degree of leak proofing of the shells studied is regarded as a feature of the particular design and not as a fixed constructional specification. The authors assess the leak proofing properties of each type of shell and establish building costs for each of them on the basis of precise plans, with the collaboration of various specialized firms. They estimate the effectiveness of the various shells from a safety standpoint, in relation to different emergency procedures, in particular release into the atmosphere through appropriate filters and decontamination of the air within the shell by recycling through batteries of filters. The paper contains a very detailed comparison of about 10 cases corresponding to various combinations of design and emergency procedure; the comparison was made using a computer programme specially established for the purpose. The results are compared with those for a reactor of the same type and power, but assembled together with the heat exchangers in a pre-stressed concrete shell. (author)

  7. A modern control room for Indian Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor (AHWR) is a next generation nuclear power plant being developed by Bhabha Atomic Research Centre, India. AHWR is a vertical, pressure tube type, heavy-water-moderated, boiling light-water-cooled, innovative reactor, relying on natural circulation for core cooling in all operating and accident conditions. In addition, it incorporates various passive systems for decay heat removal, containment cooling and isolation. In addition to the many passive safety features, AHWR has state of the art I and C architecture based on extensive use of computers and networking. In tune with the many advanced features of the reactor, a centralized modern control room has been conceived for operation and monitoring of the plant. The I and C architecture enables the implementation of a fully computerised operator friendly control room with soft Human Machine Interfaces (HMI). While doing so, safety has been given due consideration. The control and monitoring of AHWR systems are carried out from the fully computer-based operator interfaces, except safety systems, for which only monitoring is provided from soft HMI. The control of the safety systems is performed from dedicated hardwired safety system panels. Soft HMI reduces the number of individual control devices and improves their reliability. The paper briefly describes the I and C architecture adopted for the AHWR plant along with the interfaces to the main and backup control rooms. There are many issues involved while introducing soft HMI based operator interfaces for Nuclear Power Plants (NPP) compared to the conventional plants. Besides discussing the implementation issues, the paper elaborates the design considerations that have undergone in the design of various components in the main control room especially operator workstations, shift supervisor console, safety system panels and large display panels. Mainly task based displays have been adopted for the routine operator interactions of the plant

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-03-15

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

  9. Plant life management processes and practices for heavy water reactors

    International Nuclear Information System (INIS)

    In general, heavy water reactor (HWR) nuclear power plant (NPP) owners would like to keep their NPPs in service as long as they can be operated safely and economically. Their decisions are depending on essentially business model. They involve the consideration of a number of factors, such as the material condition of the plant, comparison with current safety standards, the socio-political climate and asset management/ business planning considerations. Continued plant operation, including operation beyond design life, called 'long term operation, depends, among other things, on the material condition of the plant. This is influenced significantly by the effectiveness of ageing management. Key attributes of an effective plant life management program include a focus on important systems, structure and components (SSCs) which are susceptible to ageing degradation, a balance of proactive and reactive ageing management programmes, and a team approach that ensures the co-ordination of and communication between all relevant nuclear power plant and external programmes. Most HWR NPP owners/operators use a mix of maintenance, surveillance and inspection (MSI) programs as the primary means of managing ageing. Often these programs are experienced-based and/or time-based and may not be optimised for detecting and/or managing ageing effects. From time-to-time, operational history has shown that this practice can be too reactive, as it leads to dealing with ageing effects (degradation of SSCs) after they have been detected. In many cases premature and/or undetected ageing cannot be traced back to one specific reason or an explicit error. The root cause is often a lack of communication, documentation and/or co-ordination between design, commissioning, operation or maintenance organizations. This lack of effective communication and interfacing frequently arises because, with the exception of major SSCs, such as the fuel channels or steam generators, there is a lack of explicit

  10. Calculations on heavy-water moderated and cooled natural uranium fuelled power reactors

    International Nuclear Information System (INIS)

    One of the codes that the Instituto Nacional de Investigaciones Nucleares (Mexico) has for the nuclear reactors design calculations is the LEOPARD code. This work studies the reliability of this code in reactors design calculations which component materials are the same of the heavy water moderated and cooled, natural uranium fuelled power reactors. (author)

  11. Heavy water reactor user requirement document status and path forward

    International Nuclear Information System (INIS)

    This is a Power Point presentation containing: -1. Outline; - 2. Background - Stages and the HWR-URD; - 3. Background - What do we have now?; - 4. Draft 'D0' of the HWR-URD; - 5. The 2001 April Consultancy; - 6. The Path Forward. - 7. Summary. The 2001 April consultancy addressed the following items: - 1. Status of the URD: - The document structure is substantially complete. Appendices or paragraphs need to be added on some topics; - The URD covers the overall requirements for the design of the nuclear island (NI) and interfaces to the BOP for future HWRs. It contains policies, high-level requirements and important requirements for key areas that are of interest to HWR users; - The draft focuses on horizontal-pressure-tube, heavy water moderated and cooled HWRs. When and if other types need to be considered, the TWG will identify these and direct how they are to be addressed; - The level of detail in the draft and its treatment of the international aspects of the topic are appropriate; - 2. Areas needing further consideration: - While intended for future reactors, it is recognized that regulators may wish to use the URD as a benchmark for evaluating existing or replicate reactors; - The international aspects of the URD require detailed review at each stage of its development; - The EUR and EPRI-URD have had targeted reviews by the regulators. This may be appropriate for the HWR-URD but would add 6 to 12 months to the schedule; - These items will be the focus of the AGM planned for 2002 January. The Path Forward section pinpoints the terms: - 2001 August, implying the task, incorporate comments from April Consultancy, producing draft D1; - 2002 January, implying AGM in Vienna, namely, focus on areas needing further consideration: - Ensure requirements are clearly differentiated from desirable features; - Confirm international aspects are appropriately considered; - Establish need for additional step - regulatory review; - 2002, implying revise URD to reflect

  12. Preparation Before Signature of Upgrade of Algeria Heavy Water Research Reactor Contract

    Institute of Scientific and Technical Information of China (English)

    LI; Song; ZAN; Huai-qi; XU; Qi-guo; JIA; Yu-wen

    2012-01-01

    <正>Algeria heavy water research reactor (Birine) is a multiple-purpose research reactor, which was constructed with the help of China more than 20 years ago. By request of Algeria, China will upgrade the research reactor; so as to improve the status of current reactor such as equipment ageing, shortage of spare parts, several systems do not meet requirements of current standards and criteria etc.

  13. Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors

    Directory of Open Access Journals (Sweden)

    Peel Ross

    2016-01-01

    Full Text Available This paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom. Heavy water reactors are used extensively in Canada, and are deploying within India and China, whilst the UK is considering the use of heavy water reactors to manage its plutonium inventory of 140 tonnes. The UK heavy water reactor proposal uses a mixed oxide (MOX fuel of plutonium in depleted uranium, within the enhanced CANDU-6 (EC-6 reactor. This work proposes an alternative heterogeneous fuel concept based on the same reactor and CANFLEX fuel bundle, with eight large-diameter fuel elements loaded with natural thorium oxide and 35 small-diameter fuel elements loaded with a MOX of plutonium and reprocessed uranium stocks from UK MAGNOX and AGR reactors. Indicative neutronic calculations suggest that such a fuel would be neutronically feasible. A similar MOX may alternatively be fabricated from reprocessed <5% enriched light water reactor fuel, such as the fuel of the AREVA EPR reactor, to consume newly produced plutonium from reprocessing, similar to the DUPIC (direct use of PWR fuel in CANDU process.

  14. Analysis of thorium/U-233 lattices and cores in a breeder/burner heavy water reactor

    International Nuclear Information System (INIS)

    Due to the inevitable dwindling of uranium resources, advanced fuel cycles in the current generation of reactors stand to be of great benefit in the future. Heavy water moderated reactors have much potential to make use of thorium, a currently unexploited resource. Core fuelling configurations of a Heavy Water Reactor based on the self-sufficient thorium fuel cycle were simulated using the DRAGON and DONJON reactor physics codes. Three heterogeneously fuelled reactors and one homogeneously fuelled reactor were studied. (author)

  15. Achieving safety through the design process for the heavy water new product reactor

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE) is presently completing the Conceptual Design Phase (CDP) for a heavy water new production reactor (NPR). In undertaking the development of requirements for the heavy water NPR, the DOE defined as a principal requirement that the reactor would be designed such that it would meet or exceed the level of safety and safety assurance achieved by modern commercial nuclear power plants. This paper discusses the strategy and methodology of implementing the line responsibilities for achieving safety in the design of the heavy water NPR

  16. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II. [USA

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site.

  17. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II

    International Nuclear Information System (INIS)

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site

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

    International Nuclear Information System (INIS)

    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

  19. Effect of processing on properties of thin walled calandria tubes for pressurised heavy water reactor

    Science.gov (United States)

    Kapoor, K.; Padmaprabu, C.; Ramana Rao, S. V.; Sanyal, T.; Kashyap, B. P.

    2003-02-01

    Thin walled calandria tubes for pressurised heavy water reactors are manufactured either by seam welding of Zircaloy-4 sheets or by seamless route. In the present study, the effect of processing on the critical properties such as texture, microstructure, hydriding behaviour and residual stress for both the routes as well as the mechanical anisotropy developed due to seam welding are investigated. The properties of the seam welded tube in the fusion and adjoining region are markedly different from the base material and from the seamless tube. Residual stress measurements indicate that heat affected zone (HAZ) of seam welded tubes have longitudinal tensile residual stress and the seamless tubes have uniform compressive stress along the circumference. The phase transition in the presence of residual stresses due to thermal gradient is found to modify the texture in the HAZ. The hydride orientation and mechanical anisotropy in these regions are found to be dependent on the texture of the material.

  20. Effect of processing on properties of thin walled calandria tubes for pressurised heavy water reactor

    International Nuclear Information System (INIS)

    Thin walled calandria tubes for pressurised heavy water reactors are manufactured either by seam welding of Zircaloy-4 sheets or by seamless route. In the present study, the effect of processing on the critical properties such as texture, microstructure, hydriding behaviour and residual stress for both the routes as well as the mechanical anisotropy developed due to seam welding are investigated. The properties of the seam welded tube in the fusion and adjoining region are markedly different from the base material and from the seamless tube. Residual stress measurements indicate that heat affected zone (HAZ) of seam welded tubes have longitudinal tensile residual stress and the seamless tubes have uniform compressive stress along the circumference. The phase transition in the presence of residual stresses due to thermal gradient is found to modify the texture in the HAZ. The hydride orientation and mechanical anisotropy in these regions are found to be dependent on the texture of the material

  1. A Management Strategy for the Heavy Water Reflector Cooling System of HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, H. S.; Park, Y. C.; Lim, S. P. (and others)

    2007-11-15

    Heavy water is used as the reflector and the moderator of the HANARO research reactor. After over 10 years operation since first criticality in 1995 there arose some operational issues related with the tritium. A task force team(TFT) has been operated for 1 year since September 2006 to study and deduce resolutions of the issues concerning the tritium and the degradation of heavy water in the HANARO reflector system. The TFT drew many recommendations on the hardware upgrade, tritium containing air control, heavy water quality management, waste management, and tritium measurement system upgrade.

  2. Uses of Plutonium Fuel in Pressure-Tube-Type, Heavy-Water-Moderated Thermal Reactors

    International Nuclear Information System (INIS)

    In 1962, a feasibility study was begun in the JAERI on the uses of various nuclear fuels for pressure-tube-type, heavy-water-moderated thermal reactors. This study began with analysis of the use of uranium in heavy-water-moderated thermal reactors such as the CANDU-PHW, CANDU-BLW, SGHW, EL-4, and Ref. 15, D and E lattices, which is designed in the JAERI, from the standpoint of the core design. Then, the ways of using plutonium fuel in the same types were investigated using WATCHTOWER, FLARE and VENUS codes, including: (1) direct substitution of the plutonium from light-water reactors or Magnox reactors, (2) recycle use of the plutonium from heavy-water-moderated reactors, (3) plutonium self-sustaining cycle, and (4) plutonium phoenix fuel. The following conclusions are reported: (1) In the direct substitution of plutonium, somewhat depleted plutonium is more suitable for core design than the plutonium from Magnox reactors or light-water reactors, because the increase in the initial reactivity due to large plutonium absorption cross-section must be prevented. (2) In the plutonium self-sustaining cycle, the fuel burn-up of about 15 000 ∼20000 MWd/t would be expected from natural uranium, and the positive void reactivity which always occurs in the uraniumloaded SGHW or CANDU-BLW lattices is greatly reduced, the latter property giving some margin to bum-out heat flux. (3) It may be concluded from the fuel cycle analysis that the plutonium self-sustaining cycle is equivalent to using slightly enriched uranium (about 1.0 at.%). It may be concluded that the use of plutonium in heavy-water-moderated reactors is technologically feasible and economically advantageous. (author)

  3. Study relating to the physico-chemical behaviour of heavy water in nuclear reactors

    International Nuclear Information System (INIS)

    Chemical and isotope pollution, and radiolytic decomposition are the two most important ways in which heavy water becomes degraded in nuclear reactors. Chemical pollution has led to the creation of ion exchange purification loops specially designed for reactors: the report contains a description in detail of the application of this purification method in CEA research reactors, including the analysis required, results obtained, and their interpretation. The intelligence obtained on radiolytic decomposition with the same facilities is also discussed, as well as the recombination apparatus and control equipment utilized. Finally, investigation to date in the CEA on recombination circuits for power reactors is also discussed. (author)

  4. The response time analysis of high log neutron flux rate for heavy water reactors

    International Nuclear Information System (INIS)

    The heavy water reactor such as Wolssung no. 1 has a protection/safety system named special safety system. The system has four safety systems ; shutdown no. 1, shutdown no. 2, emergency core cooling system and containment system. In this paper, the response time of high log neutron flux rate, one of the reactor trip loops of shutdown no.1/no.2, was analysed based on the description of final safety analysis report and compared to the plant measurement

  5. Design guide for category II reactors light and heavy water cooled reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-05-01

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems.

  6. Plant life management strategies for pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Ahn, Sang Bok; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    This technical report reviewed aging mechanism of the major components of CANDU 6 reactor such as pressure tubes, calandria tube, end fitting, fuel channel spacer and calandria. Furthermore, the surveillance methodology was described for monitoring and inspection of these core components. Based on the in-reactor performances data such as delayed hydride cracking, leak-before-break, enhanced deformation-creep and growth, the life management of pressure tubes was illustrated in this report. (author). 19 refs., 11 figs., 2 tabs.

  7. Pulsed Neutron Measurements on a Heavy Water Power Reactor (MZFR) at Zero Energy

    International Nuclear Information System (INIS)

    The pulsed neutron method was used for zero-power measurements in the core of a heavy water reactor. Various methods were used for the evaluation of the pulsed measurements. The so-called ''integral'' evaluation methods are based on theories published by Sjöstrand and Gozani; so far they have been applied mainly to light water reactors. These methods use not only the prompt neutron decay constant but also the information contained in the delayed neutron tails to determine the reactivity. For measurements on the heavy water reactor, however, the methods had to be modified so as to adequately take into account the time dependence of the delayed neutrons. The fraction of the delayed neutrons was calculated using a reasonable assumption for its time dependence. All the information needed could be obtained from the measurements. These methods are well suited for hand calculations to yield the reactivity with proper accuracy. An analytical procedure was applied to check the results of the integral methods. This essentially involves the exact calculation of the time dependence of the delayed neutron fraction by an iteration procedure. The results of the different evaluation methods mentioned above are compared by plotting them as functions of the D2O level and of the boron concentration. Due to the inclined control rods the flux distribution is distorted in a rather complicated manner when the rods are inserted. Therefore the time dependence of this distribution was measured for different positions of the pulsed neutron source. It was possible to find one position for which the influence of higher modes on the measurements of the shutdown reactivity was sufficiently small. Finally it is shown that the values of (δρ(H, ci)/δ(l/H2)) H = Hi and (δρ(Hi, c)/δc) c = ci (ρ reactivity, Hi critical D2O level for boron concentration c1) obtained by period measurements in the slightly supercritical state and pulsed measurements in the subcritical state are in excellent

  8. Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

    Directory of Open Access Journals (Sweden)

    Naoyuki Takaki

    2012-08-01

    Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-233U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of 233U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to 233U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

  9. Pressurized heavy water reactor fuel behaviour in power ramp conditions

    Science.gov (United States)

    Ionescu, S.; Uţă, O.; Pârvan, M.; Ohâi, D.

    2009-03-01

    In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.

  10. Graphite-moderated and heavy water-moderated spectral shift controlled reactors; Reactores de moderador solido controlados por desplazamiento espectral

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F.

    1984-07-01

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs.

  11. The status of improved pressurized heavy water reactor development - A new option for PHWR -

    International Nuclear Information System (INIS)

    Currently, the 900 MWe class Improved Pressurized Heavy Water Reactor (PHWR), which is a type of CANDU reactor based on the systems and components of operating CANDU plants, is under development. The improved PHWR has a 480 fuel channel calandria, uses 37 element natural uranium fuel bundles and has a single unit containment. Adaptation of a steel-lined containment structure and improved containment isolation systems permit a reduced exclusion area boundary (EAB) compared to the existing larger capacity CANDU reactors (Darlington, Bruce B). The improved PHWR buildings are arranged to achieve minimum spacing between reactor units. Plant safety and economy are increased through various design changes based on the operating experience of existing CANDU plants. 4 refs. (Author)

  12. Study on hydrodynamically induced dryout and post dryout important to heavy water reactors

    International Nuclear Information System (INIS)

    Recently, the safety of low pressure liquid cooled nuclear reactors has become a very important issue with reference to the operation of the heavy water reactors at Savannah River Plant. Under accident conditions such as loss-of-flow or loss-of-coolant, these reactors typically encounter unstable two-phase flow which may lead to the occurrence of dryout and subsequent fuel failure. An analytical study using the one-dimensional drift flux model was carried out to investigate the two-phase flow instability for Westinghouse Savannah River Site reactor. The analysis indicates that the first and higher order instabilities exist in the possible transient operational conditions. The instabilities are encountered at higher heat fluxes or lower flow rates. The subcooling has a stabilizing effect except at very low subcooling. An experimental loop has been designed and constructed. A study was conducted on the CHF induced by various flow instabilities. Details of this test loop are presented

  13. A review of qualitative inspection aspects of end fittings in an Indian pressurized heavy water reactor

    Directory of Open Access Journals (Sweden)

    Urva Pancholi

    2016-07-01

    Full Text Available The paper provides a summarized description of the current state of knowledge and practices used in India, in the qualitative inspection of end fittings – a key component of the fuel channel assembly of a pressurized heavy water reactor (PHWR, generally of a Canadian Deuterium Uranium (CANDU type. Further it discusses various quality inspection techniques; and the high standards and mechanical precision of the job required, to be accepted as viable nuclear reactor component. The techniques, instruments and specific data for such components mentioned here are synthesized results from primary research and knowledge available in this area, in order to produce coherent argument focused on quality control of end fittings.

  14. Fission product release phenomena during core melt accidents in metal fueled heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ellison, P G; Hyder, M L; Monson, P R; Randolph, H W [Westinghouse Savannah River Co., Aiken, SC (USA); Hagrman, D L [EG and G Idaho, Inc., Idaho Falls, ID (USA); McClure, P R; Leonard, M T [Science Applications International Corp., Albuquerque, NM (USA)

    1990-01-01

    The phenomena that determine fission product release rates from a core melting accident in a metal-fueled, heavy water reactor are described in this paper. This information is obtained from the analysis of the current metal fuel experimental data base and from the results of analytical calculations. Experimental programs in place at the Savannah River Site are described that will provide information to resolve uncertainties in the data base. The results of the experiments will be incorporated into new severe accident computer codes recently developed for this reactor design. 47 refs., 4 figs.

  15. Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

    International Nuclear Information System (INIS)

    Requests for severe accident investigations and assurance of mitigation measures have increased for operating nuclear power plants and the design of advanced nuclear power plants. Severe accident analysis investigations necessitate the analysis of the very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. The IAEA organizes coordinated research projects (CRPs) to facilitate technology development through international collaboration among Member States. The CRP on Benchmarking Severe Accident Computer Codes for HWR Applications was planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). This publication summarizes the results from the CRP participants. The CRP promoted international collaboration among Member States to improve the phenomenological understanding of severe core damage accidents and the capability to analyse them. The CRP scope included the identification and selection of a severe accident sequence, selection of appropriate geometrical and boundary conditions, conduct of benchmark analyses, comparison of the results of all code outputs, evaluation of the capabilities of computer codes to predict important severe accident phenomena, and the proposal of necessary code improvements and/or new experiments to reduce uncertainties. Seven institutes from five countries with HWRs participated in this CRP

  16. Hot functional testing of the pressurized heavy water reactor plant Atucha II with light water

    International Nuclear Information System (INIS)

    The two pressurized heavy water reactors PHWR Atucha I (designed and built by S/KWU, now AREVA), and Atucha II (designed by S/KWU and plant construction now completed by NA-SA) are owned by Nucleoelectrica Argentina S.A. (NA-SA). Atucha II was designed in the 1980'ies in parallel to the two most recent S/KWU PWR generations Prekonvoi and Konvoi. Its basic design has been updated and optimized including also backfitting of components and systems for severe accident management. The gross electric power of the plant is 745 MWe. Construction and commissioning of Atucha II has been resumed by NA-SA after a work stop in the 1990'ies and is now almost completed. Hot functional testing HFT was performed in two phases in September and October 2013 and in March and April 2014. Hot functional testing was performed with light water and the fuel assemblies loaded. The chemistry program for the HFT was derived from practices and experience gathered at other S/KWU designed PWRs during HFTs and consisted of the following main targets and requirements: (1) Low chloride and sulfate concentrations close to normal operation values specified in the VGB water chemistry guideline for power operation of PWR plants; (2) Thorough oxygen removal during heat-up and reducing conditions through N2H4 dosing; (3) High pH value (target range 1.5 to 2 ppm Li); (4) Passivation treatment of the nuclear steam supply system NSSS at temperatures of at least 260°C for a time period of at least 120 hours; (5) Zinc addition at a constant rate of 20 g Zn per day throughout the various HFT phases. Zinc dosing was begun during the first heat-up of the plant at temperatures above approx. 150°C. Daily measurement of the zinc concentration for process control was not necessary and not required due to the elaborated zinc application procedure. The main results of the chemistry program for the HFT of plant are described and evaluated in this contribution. Data shows that all chemistry targets were met

  17. Heavy water and nonproliferation

    International Nuclear Information System (INIS)

    This report begins with a historical sketch of heavy water. The report next assesses the nonproliferation implications of the use of heavy water-moderated power reactors; several different reactor types are discussed, but the focus is on the natural uranium, on-power fueled, pressure tube reactor CANDU. The need for and development of on-power fueling safeguards is discussed. Also considered is the use of heavy water in plutonium production reactors as well as the broader issue of the relative nuclear leverage that suppliers can bring to bear on countries with natural uranium-fueled reactors as compared to those using enriched designs. The final chapter reviews heavy water production methods and analyzes the difficulties involved in implementing these on both a large and a small scale. It concludes with an overview of proprietary and nonproliferation constraints on heavy water technology transfer

  18. Graphite-moderated and heavy water-moderated spectral shift controlled reactors

    International Nuclear Information System (INIS)

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs

  19. Synergistic Water-Treatment Reactors Using a TiO2-Modified Ti-Mesh Filter

    Directory of Open Access Journals (Sweden)

    Akira Fujishima

    2013-07-01

    Full Text Available The recent applications of a TiO2-modified Ti-mesh filter (TMiP™ for water purification are summarized with newly collected data including biological assays as well as sewage water treatment. The water purification reactors consist of the combination of a TMiP, a UV lamp, an excimer VUV lamp, and an ozonation unit. The water purification abilities of the reactor were evaluated by decomposition of organic contaminants, inactivation of waterborne pathogens, and treatment efficiency for sewage water. The UV-C/TMiP/O3 reactor disinfected E. coli in aqueous suspension in approximately 1 min completely, and also decreased the number of E. coli in sewage water in 15 min dramatically. The observed rate constants of 7.5 L/min and 1.3 L/min were calculated by pseudo-first-order kinetic analysis respectively. Although organic substances in sewage water were supposed to prevent the UV-C/TMiP/O3 reactor from purifying water, the reactor reduced E. coli in sewage water continuously. On the other hand, although much higher efficiencies for decomposition of organic pollutants in water were achieved in the excimer/TMiP reactor, the disinfection activity of the reactor for waterborne pathogens was not as effective as the other reactors. The difference of efficiency between organic pollutants and waterborne pathogens in the excimer/TMiP reactor may be due to the size, the structure, and the decomposition mechanism of the organic pollutants and waterborne pathogens. These results show that a suitable system assisted by synergy of photocatalysts and other technologies such as ozonation has a huge potential as a practical wastewater purification system.

  20. Experimental studies on in-bundle ECCS injection for Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    The Advanced Heavy Water Reactor (AHWR) being designed at BARC is an innovative reactor with Thorium utilization as its major objective. It has many advanced passive safety features. One such feature is passive injection of emergency coolant after postulated Loss of Coolant Accident (LOCA). A novel feature of this injection scheme is that the injection does not take place in the header/plenum as in other reactors, but directly in to the bundle. For this purpose, the fuel cluster incorporates a central water rod which communicates with the ECCS header. The water rod extends along full length of the fuel cluster. In event of LOCA in the Main Heat Transport (MHT) system, ECC water flows from the accumulator to the water rod through ECCS header. The water flows into the bundle through holes in the water rod. The AHWR fuel cluster has fuel pins arranged in three concentric rings (of 12, 18 and 24 pins) around the central rod. While it is ensured that water does reach the fuel cluster, whether it reaches the outer ring of pins is needs investigation as the pins are closely spaced (1-3 mm gap between adjacent rods). The objective of the present experiments is to determine under what conditions (ECC flow and decay heat), the ECC water is able to rewet and cool all the fuel pins. The experiments have been done in a short, instrumented fuel bundle simulating the geometry of the AHWR fuel cluster

  1. Radio-toxicity of spent fuel of the advanced heavy water reactor.

    Science.gov (United States)

    Anand, S; Singh, K D S; Sharma, V K

    2010-01-01

    The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR. PMID:19776247

  2. Analysis of transients in advanced heavy water reactor using lumped parameter models

    Energy Technology Data Exchange (ETDEWEB)

    Manmohan Pandey; Venkata Ramana Eaga; Sankar Sastry, P. [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati (India); Gupta, S.K.; Lele, H.G.; Chatterjee, B. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)

    2005-07-01

    Full text of publication follows: Analysis of transients occurring in nuclear power plants, arising from the complex interplay between core neutronics and thermal-hydraulics, is important for their operation and safety. Numerical simulations of such transients can be carried out extensively at very low computational cost by using lumped parameter mathematical models. The Advanced Heavy Water Reactor (AHWR), being developed in India, is a vertical pressure tube type reactor cooled by boiling light water under natural circulation, using thorium as fuel and heavy water as moderator. In the present work, nonlinear and linear lumped parameter dynamic models for AHWR have been developed and validated with a distributed parameter model. The nonlinear lumped model is based on point reactor kinetics equations and one-dimensional homogeneous equilibrium model of two-phase flow. The distributed model is built with RELAP5/MOD3.2 code. Various types of transients have been simulated numerically, using the lumped model as well as RELAP5. The results have been compared and parameters tuned to make the lumped model match the distributed model (RELAP5) in terms of steady state as well as dynamic behaviour. The linear model has been derived by linearizing the nonlinear model for small perturbations about the steady state. Numerical simulations of transients using the linear model have been compared with results obtained from the nonlinear model. Thus, the range of validity of the linear model has been determined. Stability characteristics of AHWR have been investigated using the lumped parameter models. (authors)

  3. Stage 1 decommissioning of the steam generating heavy water reactor - current achievements (Winfrith)

    International Nuclear Information System (INIS)

    The Steam Generating Heavy Water Reactor (SGHWR) is located at Winfrith, Dorset, UK. It was a Nuclear Power Plant rated at 100MW(e) and operated successfully from 1968 to 1990. Decommissioning is funded by the Department of Trade and Industry and is the responsibility of the United Kingdom Atomic Energy Authority, which has contracted AEA Technology to manage the project. Decommissioning is now well advanced in accordance with a cost-effective three stage strategy. This paper introduces SGHWR, outlines the economic basis of the decommissioning strategy, summarises the practical achievements and uses the containment strategy to provide an example of environmental considerations. (Author)

  4. Status of deuterium nuclear data for the simulation of heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S.; Roubtsov, D.; Rao, R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Svenne, J.P. [Univ. of Manitoba, Winnipeg, Manitoba (Canada); Winnipeg Inst. for Theoretical Physics, Winnipeg, Manitoba (Canada); Canton, L. [Inst. Nazionale de Fisica Nucleare, Sezione di Padova, Padova (Italy); Univ. di Padova, Dipartimento di Fisica, Padova (Italy); Plompen, A.J.M. [EC-JRC, Inst. for Reference Materials and Measurements, Retieseweg, Geel (Belgium); Stanoiu, M. [Horia Hulubei National Inst. for Physics and Nuclear Engineering, Magurele (Romania); Nankov, N.; Rouki, C. [EC-JRC, Inst. for Reference Materials and Measurement, Retieseweg, Geel (Belgium)

    2011-07-01

    An overview is presented of the status of the deuterium nuclear data used in reactor physics simulations of heavy water (D{sub 2}O) reactors and of ongoing activities to improve their accuracy. The main subjects having noticeable reactivity impact for critical systems involving D{sub 2}O are the degree of backscatter in D(n,n)D elastic scattering at neutron energies <3.2 MeV, the value of the elastic scattering cross section at thermal neutron energies and the adequacy of their numerical representation in evaluated nuclear data libraries. The scope includes fundamental nuclear-data measurements; three-body nuclear-theory calculations; and MCNP5 simulations of experiments involving D{sub 2}O or deuterated targets. (author)

  5. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  6. Technologies for tritium control in fission reactors moderated with heavy water

    International Nuclear Information System (INIS)

    This study was done within a program one of whose objectives was to analyze the possible strategies and technologies, to be applied to HWR at Argentine nuclear power plants, for tritium control. The high contribution of tritium to the total dose has given rise to the need by the operators and/or designers to carry out developments and improvements to try to optimize tritium control technologies. Within a tritium control program, only that one which includes the heavy water detritiation will allow to reduce the tritium concentrations at optimum levels for safety and cost-effective power plant operation. The technology chosen to be applied should depend not only on the technical feasibility but also on the analysis of economic and juncture factors such as, among others, the quantity of heavy water to be treated. It is the authors' belief that AECL tendency concerning heavy water treatment in its future reactors would be to employ the CECE technology complemented with immobilization on titanium beds, with the 'on-line' detritiation in each nuclear power plant. This would not be of immediate application since our analysis suggests that AECL would assume that the process is under development and needs to be tested. (author). 21 refs

  7. Definition and Analysis of Heavy Water Reactor Benchmarks for Testing New Wims-D Libraries

    International Nuclear Information System (INIS)

    This work is part of the IAEA-WIMS Library Update Project (WLUP). A group of heavy water reactor benchmarks have been selected for testing new WIMS-D libraries, including calculations with WIMSD5B program and the analysis of results.These benchmarks cover a wide variety of reactors and conditions, from fresh fuels to high burnup, and from natural to enriched uranium.Besides, each benchmark includes variations in lattice pitch and in coolants (normally heavy water and void).Multiplication factors with critical experimental bucklings and other parameters are calculated and compared with experimental reference values.The WIMS libraries used for the calculations were generated with basic data from JEF-2.2 Rev.3 (JEF) and ENDF/B-VI iNReleaseln 5 (E6) Results obtained with WIMS-86 (W86) library, included with WIMSD5B package, from Windfrith, UK with adjusted data, are included also, for showing the improvements obtained with the new -not adjusted- libraries.The calculations with WIMSD5B were made with two methods (input program options): PIJ (two-dimension collision probability method) and DSN (one-dimension Sn method, with homogenization of materials by ring).The general conclusions are: the library based on JEF data and the DSN meted give the best results, that in average are acceptable

  8. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    International Nuclear Information System (INIS)

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation

  9. The possible use of cermet fuel in the DIDO and PLUTO heavy-water research reactors

    International Nuclear Information System (INIS)

    International restrictions on the supply of highly enriched uranium have resulted in the requirement to fuel research reactors with a lower-enrichment uranium fuel. A study has been made of the feasibility of using low-enrichment fuels of a new type in the DIDO and PLUTO reactors. This work has been done as a contribution to the studies currently being carried out internationally on the implications of using lower-enrichment fuels in heavy-water-moderated research reactors. The uranium content of the U/Al alloy at present used cannot be increased sufficiently to maintain the requisite U235 content without undesirable effects on the physical properties of the alloy. A different type of fuel will therefore be required to maintain the desired nuclear characteristics. A possible solution to the problem is the use of a cermet (U3O8/Al) fuel material. Cermet fuel has poorer thermal conductivity than metallic fuel, and may also contain particles of the ceramic of a size that approaches the total thickness of the cermet core. We therefore have to consider both the average temperature of the centre of the fuel and whether large particles of the ceramic may be significantly hotter than the average. This paper describes a preliminary study of the feasibility of this concept from the heat-transfer and safety viewpoints. Calculations have been made for a cermet of 20%-enrichment 2.3g U/cm3, used in a high-power element in a DIDO-type reactor. To accommodate the cermet, the cladding has been reduced in thickness to 0.318mm (0.0125 in) the core increasing to 1.044mm, but the fuel geometry is otherwise unchanged. It is concluded that from the heat-transfer viewpoint there is no problem during normal operation or the maximum credible power transient in these reactors. (author). 10 refs, 6 figs, 2 tabs

  10. Fast removal of heavy metal ions and phytic acids from water using new modified chelating fiber

    Institute of Scientific and Technical Information of China (English)

    Li Xu; Jin Nan Wang; Ying Meng; Ai Min Li

    2012-01-01

    The graft copolymerization of acrylic acid (AA) onto polyethylene glycol terephthalate (PET) fiber initiated by benzoy peroxide (BPO) was carried out in heterogeneous media.Moreover,modification of the grafted PET fiber (PET-AA) was done by changing the carboxyl group into acylamino group through the reaction with dimethylamine.The modified chelating fiber (NDWJN 1) was characterized using elementary analysis,SEM and FT-IR spectroscopy.Adsorption kinetic curves indicated that NDWJN1 could fast remove heavy metal ions and phytic acids from water effectively.Furthermore,batch kinetic studies indicated that heavy metal ions adsorbed to NDWJN1 could be fitted well by both pseudo-first-order and pseudo-second-order adsorption equations,but the intra-particle diffusion plaved a dominant role in the adsorption of phvtic acids.

  11. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    International Nuclear Information System (INIS)

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative

  12. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  13. Study of the developmental status and operational features of heavy water reactors. Final report

    International Nuclear Information System (INIS)

    The Canadian Heavy Water Reactor System, CANDU, and future plans for it as reported in a number of recent Canadian papers, are briefly reviewed. This has been done recognizing several leading observers of the nuclear industry suggest CANDU may represent a promising longer term route to nuclear fuel self sufficiency. It now appears that the concept is well established with a demonstrated performance comparable to the best LWR installations, but it presently has fuel utilization efficiency which is of the same order as that of the current LWRs. It should also be noted that CANDU is designed to a different set of safety and regulatory criteria; a point having potential for at least a second order economic significance to many non-Canadian markets. The feature of particular interest, however, is that the system is indicated by the Canadian Government Corporation-Atomic Energy of Canada Ltd.- to be capable of upgrading to a thermal near breeder. As such, this is a point of substantial import, given that system hardware is already proven, of demonstrated safety, reliability and economy, and one for which some supply industry capability is now in place. Remaining areas of uncertainty in the planned CANDU program are those related to the near breeder fuel cycle with its implied fuel processing and recycling requirements, the economics of future heavy water supply, and the actual demonstration of the physics of a unity converter/near breeder. It is noted that recently a variant of this LWBR using heavy water coolant, has been proposed by U.S. workers. Thus LWBR related efforts may be a more appropriate direction for U.S. investigations than a direct focus on CANDU itself. It is recommended that further investigation of such improved conversion developments for existing LWRs be conducted; they could represent a potential addition to the present U.S. program for a 100 percent LMFBR future

  14. Residual radioactivity guidelines for the heavy water components test reactor at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B. Smith, R.; McNeil, J.

    1997-04-01

    Guidelines were developed for acceptable levels of residual radioactivity in the Heavy Water Components Test Reactor (HWCTR) facility at the conclusion of its decommissioning. Using source terms developed from data generated in a detailed characterization study, the RESRAD and RASRAD-BUILD computer codes were used to calculate derived concentration guideline levels (DCGLs) for the radionuclides that will remain in the facility. The calculated DCGLs, when compared to existing concentrations of radionuclides measured during a 1996 characterization program, indicate that no decontamination of concrete surfaces will be necessary. Also, based on the results of the calculations, activated concrete in the reactor biological shield does not have to be removed, and imbedded radioactive piping in the facility can remain in place. Viewed in another way, the results of the calculations showed that the present inventory of residual radioactivity in the facility (not including that associated with the reactor vessel and steam generators) would produce less than one millirem per year above background to a hypothetical individual on the property. The residual radioactivity is estimated to be approximately 0.04 percent of the total inventory in the facility as of March, 1997. According to the results, the only radionuclides that would produce greater than 0.0.1-millirem per year are Am-241 (0.013 mrem/yr at 300 years), C-14 (0.022 mrem/yr at 1000 years) and U-238 (0.034 mrem/yr at 6000 years). Human exposure would occur only through the groundwater pathways, that is, from water drawn from, a well on the property. The maximum exposure would be approximately one percent of the 4 millirem per year ground water exposure limit established by the U.S. Environmental Protection Agency. 11 refs., 13 figs., 15 tabs.

  15. CFD analysis of flow and temperature distribution inside the calandria of Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Passive systems are being examined for the future AHWR reactor designs. One of these systems is the passive moderator cooling system, which removes heat from the moderator in case of a Station Black Out (SBO). The heavy-water moderator gets heated due to the residual heat from the core structures and rises upward due to buoyancy. This is cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator (Calandria) pressure beyond safe limits. In this paper CFD investigations are carried out to study the temperature distributions and flow distribution inside the Calandria using a three-dimensional CFD code, OpenFoam 2.2.0. The results provide a band of operable mass flow rates which are safe for operation by virtue of prediction of hot spots in the Calandria. (author)

  16. Liquid Metal Fast Breeder Reactor Program: Argonne facilities

    Energy Technology Data Exchange (ETDEWEB)

    Stephens, S. V. [comp.

    1976-09-01

    The objective of the document is to present in one volume an overview of the Argonne National Laboratory test facilities involved in the conduct of the national LMFBR research and development program. Existing facilities and those under construction or authorized as of September 1976 are described. Each profile presents brief descriptions of the overall facility and its test area and data relating to its experimental and testing capability. The volume is divided into two sections: Argonne-East and Argonne-West. Introductory material for each section includes site and facility maps. The profiles are arranged alphabetically by title according to their respective locations at Argonne-East or Argonne-West. A glossary of acronyms and letter designations in common usage to describe organizations, reactor and test facilities, components, etc., involved in the LMFBR program is appended.

  17. Computation and measurement of calandria tube sag in pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Calandria tubes and liquid injection shutdown system nozzles in a pressurized heavy water reactor are to sag due to irradiation creep and growth during plant operation. When the sag of calandria tube becomes bigger, the calandria tube possibly comes in contact with liquid injection shutdown system tube crossing beneath the calandria tube. The contact subsequently may cause the damage on the calandria tube resulting in unpredicted outage of the plant. It is therefore necessary to check the gap between the two tubes in order to periodically confirm no contact by using a proper measure during the plant life. An ultrasonic gap measurement probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sag between both tubes in the reactor. It was found that the centerlines of calandria tubes and liquid injection shutdown system tubes can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. Based on the irradiation creep equation and the measurement data, a computer program to calculate the sags was also developed. With the computer program, the sag at the end of plant life was predicted

  18. Comparison Of The Worth Of Critical And Exponential Measurements For Heavy-Water-Moderated Reactors

    International Nuclear Information System (INIS)

    Critical and exponential experiments in general produce overlapping information on reactor lattices. Over the past ten years the Savannah River Laboratory has been operating a heavy-water critical, the PDP, and an exponential, the SE, in parallel. This paper summarizes SRL experience to give results and recommendations as to the applicability of the two kinds of facilities in different experiments. Six types of experiments are considered below: (1) Buckling measurements in uniform isotropic lattices Here Savannah River has made extensive comparisons between single-region criticals, exponentials, substitution criticals, and PCTR type measurements. The only difficulties in the exponentials seem to lie in the radial-buckling determinations. If these can be made successfully, the exponentials can offer good competition to the criticals. Material requirements are greatest for the single-region criticals, roughly comparable for the substitution criticals and exponentials, and least for the PCTR measurements. (2) Anisotropic and void effects SRL experiments with the criticals and with critical-exponential comparisons are reviewed briefly here and at greater length in a companion paper. (3) Evaluation of control systems Adequately analysed exponential experiments appear to give good results for total-worth measurements. However, for adequate study of overall flux shaping, flux tilts, etc. a full-sized critical such as the PDP is required. (4) Temperature coefficients Exponential experiments provide an excellent method for determining the temperature coefficient of buckling for uniform lattice heating. A special facility, the PSE, at Savannah River permits such measurements up to temperatures of 215°C. For non-uniform lattice heating criticals are generally preferred. (5) Mixed lattices Actual reactors rarely use the simple uniform lattices to which the exponentials basically apply. Critical experiments with mixed loadings are used at SRL both in measuring direct effects

  19. Fuzzy-like PD controller for spatial control of advanced heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Londhe, P.S., E-mail: pandurangl97@gmail.com [Research Scholar, SGGS Institute of Engineering and Technology, Vishnupuri, Nanded 431606 (India); Patre, B.M., E-mail: bmpatre@ieee.org [Department of Instrumentation Engineering, Shri Guru Gobind Singhji Institute of Engineering and Technology, Vishnupuri, Nanded 431 606 (India); Tiwari, A.P., E-mail: aptiwari@barc.gov.in [Reactor Control Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2014-07-01

    Highlights: • Highly non-linear model of AHWR is used for spatial power control. • A simple fuzzy-like PD (FZ-PD) control structure with robust rule base is developed. • Robust rule structure reduces the difficulties in design and tuning of controller. • Proposed FZ-PD structure shows robust and better transient performance. • Proposed FZ-PD controller is able to suppress spatial oscillations in AHWR. - Abstract: Spatial oscillations in the neutron flux distribution due to xenon reactivity feedback requires stringent control in large nuclear reactors, like advanced heavy water reactor (AHWR). If the spatial oscillations in the power distribution are not controlled, power density and rate of change of power at some locations in the reactor core may exceed limits of fuel failure due to ‘flux tilting’. Further, situations such as on-line refueling might cause transient variations in flux-shape from the nominal flux-shape. For analysis and control of spatial oscillations in AHWR, it is necessary to design a suitable control strategy, which will stabilize these oscillations. In this paper, a simplified scheme to design a conventional fuzzy logic controller for spatial control of AHWR is presented. This scheme known as fuzzy-like proportional derivative (FZ-PD) controller, uses robust PD (proportional derivative) type rule base. Due to robust rule base structure, tuning of scaling factors is greatly reduced. The non-linear coupled core neutronics-thermal hydraulics model of AHWR considered here represented by 90 first order differential equations. Through the dynamic simulations, it is observed that the designed FZ-PD controller is able to suppress spatial oscillations developed in AHWR and its performance is found to be robust.

  20. Measurement of internal diameter of pressure tubes in pressurized heavy water reactors using ultrasonics

    International Nuclear Information System (INIS)

    The Pressure Tube in Pressurized Heavy Water Reactors (PHWRs) undergoes dimensional changes due to the effects of creep and growth as it is subjected to high pressure and temperature, which causes Pressure Tubes to permanently increase in length and diameter and to sag because of weight of fuel and coolant (heavy water) contained in it. These dimensional changes are due to prolonged stresses under high temperature and radiation. Pressure Tube stresses are evaluated for both beginning and end of life for accounting the Pressure Tube dimensional changes that occur during its design life. At the beginning of life, the initial wall thickness and un-irradiated material properties are applied. At the end of life, Pressure Tube diameter and length increases, while wall thickness decreases. Material strength also increases during that period. The increase in Pressure Tube diameter results in squeezing of garter spring spacer between the pressure and calandria Tubes. It also causes unacceptable heat removal from the fuel due to an increased amount of primary coolant that bypasses the fuel bundles. This reduces the critical channel power at constant flow. Hence the periodic monitoring of pressure Tube diameter is important for these reasons. This is also required as per the applicable codes and standards for In-Service Inspection of PHWRs. Mechanical measurement from ID of the Tube during periodic monitoring is not practically feasible due to high radiation and inaccessibility. This necessitates the development of NDT technique using Ultrasonics for periodic in-situ measurement of ID of pressure Tubes with a BARC made remotely operated drive system called BARCIS (BARC Channel Inspection system). The development of Ultrasonic based ID measurement techniques and their actual applications in PHWRs Pressure tubes are being discussed in this paper. (author)

  1. A modified firefly algorithm applied to the nuclear reload problem of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Iona Maghali Santos de; Schirru, Roberto, E-mail: ioliveira@con.ufrj.b, E-mail: schirru@lmp.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear

    2011-07-01

    The Nuclear Reactor Reload Problem (NRRP) is an issue of great importance and concern in nuclear engineering. It is the problem related with the periodic operation of replacing part of the fuel of a nuclear reactor. Traditionally, this procedure occurs after a period of operation called a cycle, or whenever the nuclear power plant is unable to continue operating at its nominal power. Studied for more than 40 years, the NRRP still remains a challenge for many optimization techniques due to its multiple objectives concerning economics, safety and reactor physics calculations. Characteristics such as non-linearity, multimodality and high dimensionality also make the NRRP a very complex optimization problem. In broad terms, it aims at getting the best arrangement of fuel in the nuclear reactor core that leads to a maximization of the operating time. The primary goal is to design fuel loading patterns (LPs) so that the core produces the required energy output in an economical way, without violating safety limits. Since multiple feasible solutions can be obtained to this problem, judicious optimization is required in order to identify the most economical among them. In this sense, this paper presents a new contribution in this area and introduces a modified firefly algorithm (FA) to perform LPs optimization for a pressurized water reactor. Based on the original FA introduced by Xin-She Yang in 2008, the proposed methodology seems to be very promising as an optimizer to the NRRP. The experiments performed and the comparisons with some well known best performing algorithms from the literature, confirm this statement. (author)

  2. A modified firefly algorithm applied to the nuclear reload problem of a pressurized water reactor

    International Nuclear Information System (INIS)

    The Nuclear Reactor Reload Problem (NRRP) is an issue of great importance and concern in nuclear engineering. It is the problem related with the periodic operation of replacing part of the fuel of a nuclear reactor. Traditionally, this procedure occurs after a period of operation called a cycle, or whenever the nuclear power plant is unable to continue operating at its nominal power. Studied for more than 40 years, the NRRP still remains a challenge for many optimization techniques due to its multiple objectives concerning economics, safety and reactor physics calculations. Characteristics such as non-linearity, multimodality and high dimensionality also make the NRRP a very complex optimization problem. In broad terms, it aims at getting the best arrangement of fuel in the nuclear reactor core that leads to a maximization of the operating time. The primary goal is to design fuel loading patterns (LPs) so that the core produces the required energy output in an economical way, without violating safety limits. Since multiple feasible solutions can be obtained to this problem, judicious optimization is required in order to identify the most economical among them. In this sense, this paper presents a new contribution in this area and introduces a modified firefly algorithm (FA) to perform LPs optimization for a pressurized water reactor. Based on the original FA introduced by Xin-She Yang in 2008, the proposed methodology seems to be very promising as an optimizer to the NRRP. The experiments performed and the comparisons with some well known best performing algorithms from the literature, confirm this statement. (author)

  3. Multiple recycling of fuel in prototype fast breeder reactor in a closed fuel cycle with pressurized heavy-water reactor external feed

    Indian Academy of Sciences (India)

    G Pandikumar; A John Arul; P Puthiyavinayagam; P Chellapandi

    2015-10-01

    A fast breeder reactor (FBR) closed fuel cycle involves recycling of the discharged fuel, after reprocessing and refabrication, in order to utilize the unburnt fuel and the bred fissile material. Our previous study in this regard for the prototype fast breeder reactor (PFBR) indicated the possibility of multiple recycling with self-sufficiency. It was found that the change in Pu composition becomes negligible (less than 1%) after a few cycles. The core-1 Pu increases by 3% from the beginning of cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th by only 0.3%. In this work, the possibility of multiple recycling of PFBR fuel with external plutonium feed from pressurized heavy-water reactor (PHWR) is examined. Modified in-core cooling and reprocessing periods are considered. The impact of multiple recycling on PFBR core physics parameters due to the changes in the fuel composition has been brought out. Instead of separate recovery considered for the core and axial blankets in the earlier studies, combined fuel recovery is considered in this study. With these modifications and also with PHWR Pu as external feed, the study on PFBR fuel recycling is repeated. It is observed that the core-1 initial Pu inventory increases by 3.5% from cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th is only 0.35%. A comparison of the studies done with different external plutonium options viz., PHWR and PFBR radial blanket has also been made.

  4. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  5. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300 degrees C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered

  6. Indian advanced heavy water reactor for thorium utilisation and nuclear data requirements and status

    International Nuclear Information System (INIS)

    BARC is embarking on thorium utilisation program in a concerted and consistent manner to achieve all round capabilities in the entire Thorium cycle under the Advanced Heavy Water Reactor (AHWR) development program. Upgrading our nuclear data capability for thorium cycle is one of the main tasks of this program. This paper gives a brief overview of the physics design features of the AHWR. The basic starting point of the analysis has been the lattice simulation of the fuel cluster employing the WIMS-D4 code package with 1986 version of 69 group library. For the analysis of thorium cycle, the present multi group version contains the three major isotopes viz., 232Th, 233U and 233Pa. To correctly evaluate the fuel cycle we require many more isotopes of the Th burnup chain. With the help of NDS, IAEA, many other isotopes of interest in AHWR, actinides in the thorium burnup chain, burnable absorbers, etc., were generated. Some of them were added to the WIMS-D4 library and the results are discussed. The WIMS-D4 library is also being updated as part of the IAEA coordinated research project on Final Stage of WLUP with international cooperation. India is also taking part in CRP. The evaluation of AHWR lattice with this new library is presented. Some comments regarding the fission product data being used in WIMS libraries are given, which are tuned to U-Pu cycles. The measurements for 233U are rather old. Measurements in high energies are also very sparse. More attention by nuclear data community is required in this regard as well. India has also begun a modest program to assess the ADS concepts, with the aim of employing thermal reactor systems, such as AHWR. A one way coupled booster reactor concept is being analysed with available code systems and nuclear data. A brief summary of this concept is also being discussed in this paper. A general survey on the quality of the evaluated nuclear data of the major and minor isotopes of thorium cycle is also given. A major

  7. CFD Application and OpenFOAM on the 2-D Model for the Moderator System of Heavy-Water Reactors

    International Nuclear Information System (INIS)

    The flow in the complex pipeline system in a calandria tank of CANDU reactor is transported through the distribution of heat sources, which also exerts the pressure drop to the coolant flow. So the phenomena should be considered as multi-physics both in the viewpoints of heat transfer and fluid dynamics. In this study, we have modeled the calandria tank system as two-dimensional simplified one preliminarily that is yet far from the real objects, but to see the essential physics and to test the possibility of the present CFD(computational fluid dynamics) methods for the thermo-hydraulic problem in the moderator system of heavy-water reactors

  8. Trend of R&D publications in Pressurised Heavy Water Reactors: A Study using INIS and Other Databases

    OpenAIRE

    Vijai Kumar, *; Kalyane, V. L.; Prakasan, E.R.; Anil Kumar; Anil Sagar, *; Lalit Mohan

    2004-01-01

    Digital databases INIS (1970-2002), INSPEC (1969-2002), Chemical Abstracts (1977-2002), ISMEC (1973-June2002), Web of Sciences (1974-2002), and Science Citation Index (1982-2002), were used for comprehensive retrieval of bibliographic details of research publications on Pressurized Heavy Water Reactor (PHWR) research. Among the countries contributing to PHWR research, India (having 1737 papers) is the forerunner followed by Canada (1492), Romania (508) and Argentina (334). Collaboration of Ca...

  9. Ultrasonic evaluation of end cap weld joints of fuel elements of pressurized heavy water reactors using signal analysis methods

    International Nuclear Information System (INIS)

    This paper describes the application of ultrasonic digital signal analysis for the detection of fine defects of the order of 10% or lower of wall thickness (WT) of 370 microns in the resistance welded end cap-cladding tube joints of fuel elements used in Pressurised Heavy Water Reactors (PHWR s). The results obtained for the detection of such defects, have confirmed the sensitivity and reliability of this approach, and were further validated by destructive metallography. (author)

  10. Thermal-hydraulic analysis of a heavy-water reactor moderator tank using the CUPID Code

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Su Ryong; Jeong, Jae Jun [Pusan National Univ., Busan (Korea, Republic of); Kim, Hyoung Tae; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, a preliminary analysis is performed for the CANDU moderator tank. The calculation results using the basic case input showed a unrealistic, thermal stratification in the upper region, which was caused by the lack of the momentum of the cooling water from the inlet nozzle. To increase the flow momentum from the inlet nozzle, the cross-section area of each inlet nozzle was reduced by half and, then, the calculation showed very realistic results. It is clear that the modeling of the inlet nozzle affects the calculation result significantly. Further studies are needed for a realistic and efficient simulation of the flow in the Calandria tank. When the core cooling system fails to remove the decay heat from the fuel channels during a loss of coolant accident (LOCA), the pressure tube (PT) could strain to contact its surrounding Calandria tube (CT), which leads to sustained CTs dry out, finally resulting in damages to nuclear fuel. This situation can occur when the degree of the subcooling of the moderator inside the Calandria vessel is insufficient. In this regard, to estimate the local subcooling of the moderator inside the Calandria vessel is very important. However, the local temperature is measured at the inlet and outlet of the vessel only. Therefore, we need to accurately predict the local temperature inside the Calandria vessel.In this study, the thermal-hydraulic analysis of the real-scale heavy-water reactor moderator is carried out using the CUPID code. The applicability of the CUPID code to the analysis of the flow in the Calandria vessel has been assessed in the previous studies.

  11. Development of a 500 litre drum for the encapsulation of steam-generating heavy-water reactor sludge at Winfrith

    Energy Technology Data Exchange (ETDEWEB)

    Appleton, J.; Wise, M.; Staples, A.T

    2003-07-01

    Winfrith Technology Centre was once a leading UKAEA development site for nuclear technology. UKAEA's task now is to decommission the nuclear reactors and other facilities and restore the site for alternative use. On the site is the prototype steam-generating heavy-water reactor (SGHWR) that produced 100 MW(e) of electricity during its 22 year operational life. During this period the reactor produced large quantities of radioactive sludge and there are also the remains of ion exchange resins from various clean-up operations including the circuit decontamination campaigns at each annual shutdown. These sludges were directed to and stored in four external tanks and over the years there has been a steady build-up of sludge in these facilities, until 1990 when the reactor shut down permanently. Plans were made for the sludge in these tanks to be retrieved, encapsulated in drums and stored on site until a permanent national repository became available. Due to changes in circumstances following the relatively sudden closure of the SGHWR in 1990 and additional requirements from the operators of the Drigg low-level waste site, the original encapsulation plans of UKAEA had to be set aside. Following further considerations, RWE NUKEM, working in partnership with UKAEA, is now contracted to retrieve, condition and encapsulate the sludge in a new plant currently being constructed, and to export the drums to an existing refurbished on-site store. The Winfrith treated radwaste store (TRS) was constructed to store 500 1 drums of intermediate-level waste in a matrix stacked nine high. This paper describes the drum development work undertaken prior to the introduction of RWE NUKEM and completion of the revised design of drum for use within the TRS. It also briefly describes the process for which the drum is being utilised in the newly designed sludge treatment plant. The drum design has had a number of iterations from a concept that was first drafted in 1989 to the present

  12. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B.

    1997-04-01

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

  13. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    International Nuclear Information System (INIS)

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future

  14. Feasibility analysis of modified AL-6XN steel for structure component application in supercritical water-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Xinggang LI; Qingzhi YAN; Rong MA; Haoqiang WANG; Changchun GE

    2009-01-01

    Modified AL-6XN austenite steel was patterned after AL-6XN superaustenitic stainless steel by introducing microalloy elements such as zirconium and titanium in order to adapt to recrystallizing thermo-mechanical treatment and further improve crevice corrosion resistance. Modified AL-6XN exhibited comparable tensile strength, and superior plasticity and impact toughness to commercial AL-6XN steel. The effects of aging behavior on corrosion resistance and impact toughness were measured to evaluate the qualification of modified AL-6XN steel as an in-core component and cladding material in a supercritical water-cooled reactor. Attention should be paid to degradation in corrosion resistance and impact toughness after aging for 50 hours when modified AL-6XN steel is considered as one of the candidate materials for in-core components and cladding tubes in supercritical water-cooled reactors.

  15. 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)

    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

  16. 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.

  17. Study on recycle of materials and components from waste streams during decommissioning for heavy water research reactor

    International Nuclear Information System (INIS)

    The recycle of valuable materials from potential waste streams is one of important elements of waste minimization, and it can minimize the environment impact. The recycle of the arising was researched with taking the decommissioning of heavy water research reactor (HWRR) in China Institute of Atomic Energy as an example. By analyzing all the possible wastes that could generate during the decommissioning of HWRR, some amount of materials have potential values to recycle and may be used either directly or after appropriate treatment for other purposes. The research results show that in HWRR decommissioning at least tons of irons, 10 tons of aluminum and 5 tons of heavy water can be recycled by carrying out the waste minimization control measures (eg. waste classification and waste stream segregation), adopting appropriate decontamination technologies, and performing the requirements of clearance. (authors)

  18. Damage evaluation of 500 MWe Indian pressurized heavy water reactor nuclear containment for air craft impact

    International Nuclear Information System (INIS)

    Non-linear transient dynamic analysis of 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) nuclear containment has been carried out for the impact of Boeing and Airbus category of aircraft operated in India. The impulsive load time history is generated based on the momentum transfer of the crushable aircraft (soft missiles) of Boeing and Airbus families on the containment structure. The case studies include the analyses of outer containment wall (OCW) single model and the combined model with outer and inner containment wall (ICW) for impulsive loading due to aircraft impact. Initially the load is applied on OCW single model and subsequently the load is transferred to ICW after the local perforation of the OCW is noticed in the transient simulation. In the first stage of the analysis it is demonstrated that the OCW would suffer local perforation with a peak local deformation of 117 mm for impact due to B707-320 and 196 mm due to impact of A300B4 without loss of the overall integrity. However, this first barrier (OCW) cannot absorb the full impulsive load. In the second stage of the analysis of the combined model, the ICW is subjected to lower impulse duration as the load is transferred after 0.19 sec for B707-320 and 0.24 sec for A300B4 due to the local perforation of OCW. This results in the local deformation of approx. 115 mm for B707-320 and 124 mm for A300B4 in ICW and together both the structures (OCW and ICW) are capable of absorbing the full impulsive load. The analysis methodology evolved in the present work would be useful for studying the behaviour of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due commercial aircraft operated in India. (author)

  19. Boiling Water Reactor Loading Pattern Optimization Using Simple Linear Perturbation and Modified Tabu Search Methods

    International Nuclear Information System (INIS)

    An automated system for designing a loading pattern (LP) for boiling water reactors (BWRs) given a reference LP and control rod (CR) sequence has been developed. This system employs the advanced nodal code SIMULATE-3 and a BWR LP optimization code FINELOAD-3, which uses a simple linear perturbation method and a modified Tabu search method to select potential optimized LP candidates. Both of these unique methods of FINELOAD-3 were developed to achieve an effective BWR LP optimization strategy and to have high computational efficiency. FINELOAD-3 also adjusts deep CR positions to compensate for the core reactivity deviation caused by fuel shuffling. The objective function is to maximize the end-of-cycle core reactivity while satisfying the specified thermal margins and cold shutdown margin constraints. This optimization system realized the practical application for real BWR LP design. Computer time needed to obtain an optimized LP for a typical BWR/5 octant core with 15 depletion steps is ∼4 h using an engineering workstation. This system was extensively tested for real BWR reload core designs and showed that the developed LPs using this system are equivalent or better than the manually optimized LPs

  20. Some questions on nuclear safety of heavy-water power reactor operating in self-sufficient thorium cycle

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.

    2008-01-01

    Full Text Available In this paper the comparative calculations of the void coefficient have been made for different types of channel reactors for the coolant density interval 0.8-0.01 g/cm3. These results demonstrate the following. In heavy-water channel reactors, the replacement of D2O coolant by H2O, ensuring significant economic advantage, leads to the essential reducing of nuclear safety of an installation. The comparison of different reactors by the void coefficient demonstrates that at the dehydration of channels the reactivity increase is minimal for HWPR(Th, operating in the self-sufficient mode. The reduction of coolant density in channels in most cases is accompanied by the increase of power and temperatures of fuel assemblies. The calculations show that the reduction of reactivity due to Doppler effect can compensate the effect of dehydration of a channel. However, the result depends on the time dependency of heat-hydraulic processes, occurring in reactor channels in the specific accident. The result obtained in the paper confirms that nuclear safety of HWPR(Th lies on the same level as nuclear safety of CANDU type reactors approved in practice.

  1. Preliminary study of the tight lattice pressured heavy water reactor loaded with Pu/U and Th/U mixed fuels

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    To improve nuclear fuel utilization efficiency and prolong fuel cycle burn-up, a tight pitch lattice pressured heavy water reactor was investigated as an alternative of next generation of power reactors. It is shown that the high conversion ratio and negative coolant void reactivity coefficient are challenges in the reactor core physics designs. Various techniques were proposed to solve these problems. In this work, a tight pitch lattice and mixed fuel assemblies pressured heavy water reactor concept was investigated. By utilizing numerical simulation technique, it is demonstrated that reactor core mixed with Pu/U and Th/U assemblies can achieve high conversion ratio (0.98), long burn-up (60 GWD/t) and negative void reactivity coefficients.

  2. Nonlinear stability analysis of a reduced order model of nuclear reactors: A parametric study relevant to the advanced heavy water reactor

    International Nuclear Information System (INIS)

    Research highlights: → We model power oscillations in boiling water reactors using a lumped parameter model. → The nature and amplitudes of oscillations is obtained using a nonlinear analysis. → The method of multiple scales has been used for the analytical treatment. → Fuel temperature coefficient of reactivity determines the nature of oscillations. → The presented systematic method of analysis useful for reduced order reactor models. - Abstract: In this paper, we perform a parametric study of the nonlinear dynamics of a reduced order model for boiling water reactors (BWR) near the Hopf bifurcation point using the method of multiple scales (MMS). Analysis has been performed for general values of the parameters, but the results are demonstrated for parameter values of the model corresponding to the advanced heavy water reactor (AHWR). The neutronics of the AHWR is modeled using point reactor kinetic equations while a one-node lumped parameter model is assumed both for the fuel and the coolant for modeling the thermal-hydraulics. Nonlinearities in the heat transfer process are ignored and attention is focused on the nonlinearity introduced by the reactivity feedback. It is found that the steady-state operation of the AHWR mathematical model looses stability via. a Hopf bifurcation resulting in power oscillations as some typical bifurcation parameter like the void coefficient of reactivity is varied. The bifurcation is found to be subcritical for the parameter values corresponding to the AHWR. However, with a decrease in the fuel temperature coefficient of reactivity the bifurcation turns to supercritical implying global stability of the steady state operation in the linear stability regime. Moreover slight intrusion into the instability regime results in small-amplitude limit cycles leaving the possibility of retracting back to stable operation.

  3. Effect of flooding of annulus space between CT and PT with light water coolant and heavy water moderator on AHWR reactor physics parameters

    International Nuclear Information System (INIS)

    In AHWR lattice, the pressure tube (PT) contains light water coolant which carries away heat generated in the fuel pins. The pressure tube (PT) and calandria tube (CT) are separated by air (density=0.0014 g/cc) of wall thickness 1.79 cm. Air between pressure tube and calandria tube acts as insulator and minimize the heat transfer from coolant to moderator which is outside the calandria tube. In case of flooding or under any unforeseeable circumstances, the air gap between the coolant tube and calandria tube may be filled with the light water coolant or heavy water moderator. This paper gives the details of effect of filling the annulus space between CT and PT with light water or heavy water moderator on reactor physics parameters. (author)

  4. Removal of heavy metals from water by zeolite mineral chemically modified. Mercury as a particular case

    International Nuclear Information System (INIS)

    Research works on the removal of mercury from water by zeolite minerals show that a small quantity of this element is sorbed. In this work the mercury sorption from aqueous solutions in the presence and absence of Cu(l l), Ni(l l) and/or Zn(l l) by a Mexican zeolite mineral, natural and modified by cisteaminium chloride or cistaminium dichloride, was investigated in acidic p H. The zeolite minerals were characterized by X- Ray diffraction Ftir, scanning electron microscopy and semiquantitative elemental analysis (EDS), surface area analysis (BET) and thermogravimetric analysis (TGA). Mercury from aqueous solutions was quantified by Atomic absorption spectroscopy. The amount of sulphur on the zeolite samples treated with Na CI and modified with cisteaminium chloride (0.375 mmol/g) or cistaminium dichloride(0.475 mmol/g) was found to be higher than that of the zeolite minerals modified with cisteaminium chloride and cistaminium dichloride without treating them with Na CI. The amount of sulphur on the zeolite minerals modified with thiourea was the lowest. The diffusion coefficients and sorption isotherms for mercury were determined in the natural, treated with Na CI and, treated with Na CI and then modified with the cisteaminium chloride or cistaminium dichloride zeolite samples. The retention of mercury was the highest for the zeolite minerals treated Na CI and then modified with cisteaminium chloride or cistaminium dichloride, with adsorption capacity of 0.0511 and 0.0525 mmol Hg/g, respectively. In this research work, it was found that the retention of mercury by the modified minerals was not affected by the presence of Cu (Il), Zn(l l) y Ni (I l) under the experimental conditions. (Author)

  5. Numerical analysis on the calandria tubes in the moderator of a heavy water reactor using OpenFOAM and other codes

    International Nuclear Information System (INIS)

    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors. (authors)

  6. Numerical Analysis on the Calandria Tubes in the Moderator of a Heavy Water Reactor Using OpenFOAM and Other Codes

    Science.gov (United States)

    Chang, Se-Myong; Kim, Hyoung Tae

    2014-06-01

    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors.

  7. Severe accident analysis to verify the effectiveness of severe accident management guidelines for large pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gokhale, O.S., E-mail: onkarsg@barc.gov.in; Mukhopadhyay, D., E-mail: dmukho@barc.gov.in; Lele, H.G., E-mail: hglele@barc.gov.in; Singh, R.K., E-mail: rksingh@barc.gov.in

    2014-10-15

    Highlights: • The progression of severe accident initiated from high pressure scenario of station black out has been analyzed using RELAP5/SCDAP. • The effectiveness of SAMG actions prescribed has been established through analysis. • The time margin available to invoke the SAMG action has been specified. - Abstract: The pressurized heavy water reactor (PHWR) contains both inherent and engineered safety features that help the reactor become resistant to severe accident and its consequences. However in case of a low frequency severe accident, despite the safety features, procedural action should be in place to mitigate the accident progression. Severe accident analysis of such low frequency event provides insight into the accident progression and basis to develop the severe accident management guidelines (SAMG). Since the order of uncertainty in the progression path of severe accident is very high, it is necessary to study the consequences of the SAMG actions prescribed. The paper discusses severe accident analysis for large PHWRs for multiple failure transients involving a high pressure scenario (initiation event like SBO with loss of emergency core cooling system and loss of moderator cooling). SAMG actions prescribed for such a scenario include water injection into steam generator, calandria vessel or calandria vault at different stages of accident. The effectiveness of SAMG actions prescribed has been investigated. It is found that there is sufficient time margin available to the operator to execute these SAMG actions and the progression of severe accident is arrested in all the three cases.

  8. Conceptual design of a large heavy water reactor for US siting

    International Nuclear Information System (INIS)

    Information on the PHWR type reactor is presented concerning design characteristics; fuel management and resource utilization; economic evaluations; safety, licensing, and environmental impact; and commercial introduction

  9. Conceptual design of a large heavy water reactor for US siting

    Energy Technology Data Exchange (ETDEWEB)

    Shapiro, N. L.; Jesick, J. F.; Molin, A. T.; Daniel, J. A.

    1979-09-01

    Information on the PHWR type reactor is presented concerning design characteristics; fuel management and resource utilization; economic evaluations; safety, licensing, and environmental impact; and commercial introduction.

  10. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, N.J.

    1998-10-21

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

  11. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    International Nuclear Information System (INIS)

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints

  12. Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials

    International Nuclear Information System (INIS)

    The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab

  13. Electroanalysis of copper as a heavy metal pollutant in water using cobalt oxide modified exfoliated graphite electrode

    Science.gov (United States)

    Ndlovu, T.; Arotiba, O. A.; Sampath, S.; Krause, R. W.; Mamba, B. B.

    Copper is one of the heavy metals that have been recognized as essential for living organisms in trace amounts as a cofactor for crucial enzymes. However, excess amount of this trace element can have serious health effects. It is therefore important to monitor Cu in drinking water as it can easily be overlooked due to its biological functions. An electrochemical technique using re-compressed exfoliated graphite modified with cobalt oxide nanoparticles was evaluated as an electrochemical sensor for the detection of Cu2+ in spiked water samples. The analysis involved an accumulation step at -500 mV while stirring followed by square wave-anodic stripping voltammetry (SW-ASV). The accumulation step resulted in the reduction of Cu2+ ions in solution onto the electrode surface which were subsequently stripped off on the second step resulting in an analytical current signal. The electrodeposition time and potential were first optimised and the best conditions were used to get a detection limit of 94 μg L-1. This sensor was used for Cu analysis in real water samples using standard addition method with percentage recoveries of between 99% and 101%.

  14. Radiological consequence analyses of loss of coolant accidents of various break sizes of Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    For any advanced technology, it is essential to ensure that the consequences associated with the accident sequences arising, if any, from the operation of the plant are as low as possible and certainly below the guidelines/limits set by the regulatory bodies. Nuclear power is no exception to this. In this paper consequences of the events arising from Loss of Coolant Accident (LOCA) sequences in Pressurized Heavy Water Reactor (PHWR), are analysed. The sequences correspond to different break sizes of LOCA followed by the operation or otherwise of Emergency Core Cooling System (ECCS). Operation or otherwise of the containment safety systems has also been considered. It has been found that there are no releases to the environment when ECCS is available. The releases, when ECCS is not available, arise from the slack and the ground. The radionuclides considered include noble gases, iodine, and cesium. The hourly meteorological parameters (wind speed, wind direction, precipitation and stability category), considered for this study, correspond to those of Kakrapar site. The consequences evaluated are the thyroid dose and the bone marrow dose received by a person located at various distances from the release point. Isodose curves are generated. From these evaluations, it has been found that the doses are very low. The complementary cumulative frequency distributions (CCFD) for thyroid and bone marrow doses have also been presented for the cases analysed. (author)

  15. Heavy water reactors: Status and projected development. Part I. Final draft of a report to be published in the IAEA technical reports series. Working material

    International Nuclear Information System (INIS)

    In 1996, the 40th General Conference of the IAEA approved the establishment of a new International Working Group (IWG) on Advanced Technologies for Heavy Water Reactors (HWR). At its first meeting, held in June 1997, the IWG-HWR advised the Agency to prepare a TECDOC to present: a) the status of HWR advanced technology in the areas of economics, safety and fuel cycle flexibility and sustainable development; and b) the advanced technology developments needed in the following two decades to achieve the vision of the advanced HWR. The IAEA convened two consultancies and two Advisory Group Meetings to prepare the TECDOC. One of the consultancies was on 'Fuel Cycle Flexibility and Sustainable Development'; the second was on 'Passive Safety Features of HWRs Status and Projected Advances'. The members of the IWG-HWR collectively agreed on the essential features that the development of HWRs must emphasize. These 'drivers' are: improved economics: the fundamental requirement for all successful high technology developments to advance, is real economic improvements, consistent with improved quality; enhanced safety: to meet increasingly stringent requirements to satisfy the regulatory authorities, the public and the operators, an evolutionary safety path will be followed, incorporating advanced passive safety concepts where it is feasible and sensible to do so; sustainable development: the high neutron economy of HWRs results in a reactor that can burn natural uranium at high utilization, utilize spent fuel from other reactor types, and, through various recycle strategies including use of thorium, extend fissile fuel resources into the indefinite future. The objectives of this document are: to present the status of HWR technology; to document the safety characteristics of current HWR designs and the potential enhancements; to present a 'vision' of the long-term development of the HWR for use into the next century as an electricity source that is sustainable and flexible and

  16. Heavy water reactors: Status and projected development. Part II. Final draft of a report to be published in the IAEA technical reports series. Working material

    International Nuclear Information System (INIS)

    In 1996, the 40th General Conference of the IAEA approved the establishment of a new International Working Group (IWG) on Advanced Technologies for Heavy Water Reactors (HWR). At its first meeting, held in June 1997, the IWG-HWR advised the Agency to prepare a TECDOC to present: a) the status of HWR advanced technology in the areas of economics, safety and fuel cycle flexibility and sustainable development; and b) the advanced technology developments needed in the following two decades to achieve the vision of the advanced HWR. The IAEA convened two consultancies and two Advisory Group Meetings to prepare the TECDOC. One of the consultancies was on 'Fuel Cycle Flexibility and Sustainable Development'; the second was on 'Passive Safety Features of HWRs Status and Projected Advances'. The members of the IWG-HWR collectively agreed on the essential features that the development of HWRs must emphasize. These 'drivers' are: improved economics: the fundamental requirement for all successful high technology developments to advance, is real economic improvements, consistent with improved quality; enhanced safety: to meet increasingly stringent requirements to satisfy the regulatory authorities, the public and the operators, an evolutionary safety path will be followed, incorporating advanced passive safety concepts where it is feasible and sensible to do so; sustainable development: the high neutron economy of HWRs results in a reactor that can burn natural uranium at high utilization, utilize spent fuel from other reactor types, and, through various recycle strategies including use of thorium, extend fissile fuel resources into the indefinite future. The objectives of this document are: to present the status of HWR technology; to document the safety characteristics of current HWR designs and the potential enhancements; to present a 'vision' of the long-term development of the HWR for use into the next century as an electricity source that is sustainable and flexible and

  17. Optimization of U–Th fuel in heavy water moderated thermal breeder reactors using multivariate regression analysis and genetic algorithms

    International Nuclear Information System (INIS)

    Highlights: • A new method useful for the parametric analysis and optimization of reactor core designs. • This uses the strengths of genetic algorithms (GA), and regression splines. • The method is applied to the core fuel pin cell of a PHWR design. • Tools like java, R, and codes like Serpent, Matlab are used in this research. - Abstract: An analysis and optimization of a set of neutronics parameters of a thorium-fueled pressurized heavy water reactor core fuel has been performed. The analysis covers a detailed pin-cell analysis of a seed-blanket configuration, where the seed is composed of natural uranium, and the blanket is composed of thorium. Genetic algorithms (GA) is used to optimize the input parameters to meet a specific set of objectives related to: infinite multiplication factor, initial breeding ratio, and specific nuclide’s effective microscopic cross-section. The core input parameters are the pitch-to-diameter ratio, and blanket material composition. Recursive partitioning of decision trees (rpart) multivariate regression model is used to perform a predictive analysis of the samples generated from the GA module. Reactor designs are usually complex and a simulation needs a significantly large amount time to execute, hence implementation of GA or any other global optimization techniques is not feasible, therefore we present a new method of using rpart in conjunction with GA. Due to using rpart, we do not necessarily need to run the neutronics simulation for all the inputs generated from the GA module rather, run the simulations for a predefined set of inputs, build a regression fit to the input and the output parameters, and then use this fit to predict the output parameters for the inputs generated by GA. The rpart model is implemented as a library using R programming language. The results suggest that the initial breeding ratio tends to increase due to a harder neutron spectrum, however a softer neutron spectrum is desired to limit the

  18. Parametric study on effect of break size during LOCA on thermal hydraulic conditions in an indian pressurized heavy water reactor (220 MWe)

    Energy Technology Data Exchange (ETDEWEB)

    Rao, G.S.; Gupta, S.K.; Raj, V.V. [Bhabha Atomic Research Centre, Mumbai (India)

    1999-07-01

    Loss Of Coolant Accident (LOCA) in a Pressurized Heavy Water Reactor (PHWR) leads to coolant expulsion in a primary heat transport system resulting in depressurization and possible core voiding. This results in deterioration of cooling conditions in reactor channels and increase in power before reactor shutdown, leading to higher fuel temperatures. Coolant expulsion rates during LOCA are dictated by critical flow conditions governed by initial plant conditions prior to the accident, break geometry, location of break, etc. In addition the PHWRs have positive void-coefficient of reactivity for coolant resulting in reactor power rise in earlier part of LOCA, when the stored heat of the fuel has yet not been removed. If, in addition, heat transfer to the coolant drops sharply very high fuel surface temperatures are expected. The paper describes analyses carried out for three different break sizes. (author)

  19. Advanced eddy current technique for measurement of annular gap between pressure tube and calandria tube in Indian Pressurized Heavy Water Reactors (PHWRs)

    International Nuclear Information System (INIS)

    In Indian Pressurised Heavy Water Reactors (PHWRs), the PT (pressure tube) is designed to be nominally concentric with the encircling CT (calandria tube). Due to various factors PT becomes eccentric with respect to CT over the life of reactor. If this becomes excessive, hot PT will come in contact with cold CT. Such a cold spot could act as potential location for initiating blister formation and premature failure of PT. Hence it is important to periodically measure annular gap between PT and CT. An advanced eddy current technique has been successfully developed and incorporated in BARCIS (BARC Channel Inspection System) for measurement of PT-CT gap. (author). 4 refs., 3 figs

  20. Estimation of large early release frequency for Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Level 2 probabilistic safety assessment (PSA) examines severe accidents through a combination of probabilistic and deterministic approaches, in order to determine the release of radionuclides from containment, including the physical processes that are involved in the loss of structural integrity of the reactor core. The probabilistic part focuses on the reliability evaluation of containment systems and the deterministic part focuses on the analysis of the physical processes of an accident (timing and magnitude of radioactivity release), and the response of the containment. The important tasks involved are: (i) grouping and categorization of accident sequences into plant damage states (PDSs) (ii) development of a containment event trees (CETs) (iii) development of CET top event definitions and quantification of failure probabilities, and (iv) assigning the release categories and estimation of large early release frequency (LERF). LERF is defined as the frequency of those accidents leading to rapid, unmitigated release of airborne fission products from the containment to the environment occurring before the effective implementation of offsite emergency response and protective actions such that there is the potential for early health effects. Such accidents generally include unscrubbed releases associated with early containment failure shortly after vessel breach, containment bypass events, and loss of containment isolation. Preliminary assessment of LERF, based on release categorization from qualitative expert judgement, has been carried out and the estimated LERF is found to be 1e-13/yr. The dominant contributors are: (a) LB LOCA with the failure of prompt shutdown coupled with containment isolation failure, (b) containment bypass event from main steam line break outside containment coupled with failure of main steam isolation valves, and (c) LB LOCA with complete failure of emergency core cooling system (ECCS) and loss of moderator cooling

  1. Isotopic effect in the radiolytic deuterium production in PHWR (pressurized heavy water reactors)

    International Nuclear Information System (INIS)

    The isotopic concentration factor α = (H atoms/D atoms)gas/(H atoms/D atoms)liquid was determined in the deuterium gas dissolved in the primary system of Atucha I Nuclear Station (CNA I) and in the cover gas of the moderator and feed water tank of the primary system in Embalse Nuclear Station (CNE). The applied gas chromatographic method allowed the determination of D2, HD and H2 in the samples. The following α values were found: 3.5 ± 1.3 for the D2 dissolved in the primary system of CNA I, and 15 ≤ 2 and 88 ± 58 for the cover gases of the feed water tank and the moderator of CNE respectively. A number of possible factors causing the changes in α were analyzed. (Author)

  2. Breeding of 233U in the thorium–uranium fuel cycle in VVER reactors using heavy water

    International Nuclear Information System (INIS)

    A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the 233U–232Th oxide fuel of water-moderated reactors with variable water composition (D2O, H2O) that ensures breeding of the 233U and 235U isotopes. The method is comparatively simple to implement

  3. Breeding of {sup 233}U in the thorium–uranium fuel cycle in VVER reactors using heavy water

    Energy Technology Data Exchange (ETDEWEB)

    Marshalkin, V. E., E-mail: marshalkin@vniief.ru; Povyshev, V. M. [Russian Federal Nuclear Center All-Russian Research Institute of Experimental Physics (VNIIEF) (Russian Federation)

    2015-12-15

    A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the {sup 233}U–{sup 232}Th oxide fuel of water-moderated reactors with variable water composition (D{sub 2}O, H{sub 2}O) that ensures breeding of the {sup 233}U and {sup 235}U isotopes. The method is comparatively simple to implement.

  4. Breeding of 233U in the thorium-uranium fuel cycle in VVER reactors using heavy water

    Science.gov (United States)

    Marshalkin, V. E.; Povyshev, V. M.

    2015-12-01

    A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the 233U-232Th oxide fuel of water-moderated reactors with variable water composition (D2O, H2O) that ensures breeding of the 233U and 235U isotopes. The method is comparatively simple to implement.

  5. An Early History of Heavy Water

    CERN Document Server

    Waltham, C

    2002-01-01

    Since 1945 Canada has had a nuclear power industry based on reactor design which uses natural uranium and heavy water. The tortuous and improbable sequence of events which led to this situation is examined.

  6. Remove of Heavy Metals(Cu2+, Pb2+, Zn2+ and Cd2+) in Water Through Modified Diatomite

    Institute of Scientific and Technical Information of China (English)

    LIU Jia; WANG Hong-liang; L(U) Chun-xin; LIU Han-fei; GUO Zhi-xin; KANG Chun-li

    2013-01-01

    At present,heavy metal pollution has become a major environmental problem,influencing the survival of human and other creatures.Developing heavy metal ion adsorbents,which are environment friendly,of low cost and good performance is one of the effective means to solve heavy metal pollution.Rich low-grade diatomite mineral,as raw material,was modified with calcium carbonate to improve the adsorption properties of diatomite for heavy metal ions.The effects of dosage(m),pH,adsorbing time(t),temperature(T) and concentration of adsorbent(c) on the removal rate of heavy metal ions(Cu2+,Pb2+,Zn2+ and Cd2+) were studied.The results show that under the suitable conditions,T=15 ℃,c=1.5 mmol/L,m=10 g/L,t=60 min,in a weak acid environment,the removal effect of the four kinds of metal ions by modified diatomite is the best.

  7. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes: Flaw detection and characterization [Phase 1

    International Nuclear Information System (INIS)

    Nuclear power plants with heavy water reactors (HWRs) comprise nine percent of today's operating nuclear units, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems are an important factor in assuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Programme (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the frame of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of the IAEA's project on advanced technologies for HWRs. The objective of the CRP was to inter-compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP, participants have investigated the capability of different techniques to detect and characterize flaws. During the second phase of this CRP, participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in Zirconium alloys. The intent was to identify the most effective pressure tube inspection and diagnostic methods, and to identify further development needs. The organizations that have participated in this CRP are: - The Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL); Chalk River Laboratories (CRL), Canada; - The Research Institute of Nuclear Power Operations (RINPO), China National Nuclear Corporation (CNNC), China; - Bhabha Atomic Research Centre (BARC), India; - The Korea Electric Power Research Institute (KEPRI), Republic of Korea; - The Korea Atomic Energy

  8. Evaluation of endcap welds in thin walled fuel elements of pressurised heavy water reactor by ultrasonic testing

    International Nuclear Information System (INIS)

    In the pressurised heavy water reactor systems of India, the fuel is encapsulated in thin-walled tubes (0.342 mm) closed with endcaps by resistance welding. The integrity of these fuel elements should be such that no fission gas leakage takes place during reactor operation. The quality control of the endcap welds needed to satisfy this requirement includes helium leak test and destructive metallographic test (on sample basis). This paper discusses the feasibility study that has been carried out in the author's laboratory to develop an immersion ultrasonic test method for evaluating the integrity of the endcap weld region. Through holes of various sizes (0.15mm, 0.2mm, 0.4mm diameter and 0.185mm and 0.342mm deep) were machined by spark erosion machining at the weld joints to simulate defects of various sizes. Line focussed probe of 10 MHz frequency was used for the testing. It was possible to detect clearly all the machined holes. Based on the above standardised procedure, further testing was done on endcap welds which were rejected during fabrication on account of showing leak rate of 3 x 10-6 std. c.c/sec. or more during helium leak test. Though it was possible to get echoes from the natural defects in the rejected tubes with echo amplitude of 70%, the signal was accompanied by the geometrical reflection (noise) giving an amplitude of 20% from the weld region, giving rise to the problem of resolving the defect indication from the geometric indications. Therefore, signal analysis approach was adopted. The signal obtained from the weld zone were subjected to various analysis procedures like a) autopower spectrum, b) total energy content and c) demodulated auto correlation function. It was possible by all the three methods to differentiate the defect signal from those due to weld geometry or due to noise. Subsequently, metallography was carried out to characterise the type of defects observed during the ultrasonic testing. (author). 4 figs

  9. Heavy water lattices: Second panel report

    International Nuclear Information System (INIS)

    The panel was attended by prominent physicists from most of the laboratories engaged in the field of heavy water lattices throughout the world. The participants presented written contributions and status reports describing the past history and plans for further development of heavy-water reactors. Valuable discussions took place, during which recommendations for future work were formulated. Refs, figs, tabs

  10. Characterization of potassium hydroxide (KOH) modified hydrochars from different feedstocks for enhanced removal of heavy metals from water.

    Science.gov (United States)

    Sun, Kejing; Tang, Jingchun; Gong, Yanyan; Zhang, Hairong

    2015-11-01

    Hydrochars produced from different feedstocks (sawdust, wheat straw, and corn stalk) via hydrothermal carbonization (HTC) and KOH modification were used as alternative adsorbents for aqueous heavy metals remediation. The chemical and physical properties of the hydrochars and KOH-treated hydrochars were characterized, and the ability of hydrochars for removal of heavy metals from aqueous solutions as a function of reaction time, pH, and initial contaminant concentration was tested. The results showed that KOH modification of hydrochars might have increased the aromatic and oxygen-containing functional groups, such as carboxyl groups, resulting in about 2-3 times increase of cadmium sorption capacity (30.40-40.78 mg/g) compared to that of unmodified hydrochars (13.92-14.52 mg/g). The sorption ability among different feedstocks after modification was as the following: sawdust > wheat straw > corn stack. Cadmium sorption kinetics on modified hydrochars could be interpreted with a pseudo-second order, and sorption isotherm was simulated with Langmuir adsorption model. High cadmium uptake on modified hydrochars was observed over the pH range of 4.0-8.0, while for other heavy metals (Pb(2+), Cu(2+), and Zn(2+)) the range was 4.0-6.0. In a multi-metal system, the sorption capacity of heavy metals by modified hydrochars was also higher than that by unmodified ones and followed the order of Pb(II) > Cu(II) > Cd(II) > Zn(II). The results suggest that KOH-modified hydrochars can be used as a low cost, environmental-friendly, and effective adsorbent for heavy metal removal from aqueous solutions.

  11. Design and tuning of a Decentralized Fuzzy Logic Controller for a MIMO type Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Highlights: • A 14 inference modules based DFLC is designed for 70th order MIMO PHWR system. • Auto tuning of DFLC for PHWR is performed using NMA. • A novel approach is presented to overcome the shortcomings of NMA in tuning the DFLC. • The optimally tuned DFLC is evaluated for robustness and reference tracking capabilities. - Abstract: A Pressurized Heavy Water Reactor (PHWR) is a highly complex and unstable system. Designing a safe, reliable and robust controller with good performance for such a large and complex system is an important control engineering problem. In this work, a Decentralized Fuzzy Logic Controller (DFLC) with 140 input and 70 output membership functions, is designed for a 70th order Multi-Input Multi-Output (MIMO) type PHWR. In order to obtain high performance of the controller, it needs to be tuned optimally, however, it is very challenging task to optimally tune the DFLC with such a large membership functions. Moreover, PHWR is a coupled system which imposes additional limitation in tuning the controller since the output of one PHWR’s zone affects the outputs of other zones. In this work, an application of Nelder–Mead Algorithm (NMA) is presented for auto tuning the DFLC. The NMA performance depends upon objective function and initial points given to the NMA at the start of the tuning process. A novel method for selecting the optimal objective function and initial points for the NMA is also proposed since their selection is another complicated process. Although several objective functions have been proposed by the researchers for use with NMA, this work focuses five common indices (IAE, ISE, ITAE, ITSE and ISTE) as objective functions, which are simple and system independent. Finally, the optimally tuned high-performance DFLC is applied to the PHWR and evaluated by simulating different scenarios. The simulation results show that the controller is efficient, fast and robust and ensures the safety and reliability of the PHWR

  12. Experimental determination of thermal contact conductance between pressure and calandria tubes of Indian pressurised heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dureja, A.K., E-mail: akdureja@barc.gov.in [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Pawaskar, D.N.; Seshu, P. [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Mumbai (India); Sinha, S.K. [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Sinha, R.K. [Department of Atomic Energy, OYC, Near Gateway of India, Mumbai (India)

    2015-04-01

    Highlights: • We established an experimental facility to measure thermal contact conductance between disc shaped specimens. • We measured thermal contact conductance between Zr-2.5Nb alloy pressure tube (PT) material and Zr-4 calandria tube (CT) material. • We concluded that thermal contact conductance is a linear function of contact pressure for interface of PT and CT up to 10 MPa contact pressure. • We concluded that thermal contact conductance is a weak function of interface temperature. - Abstract: Thermal contact conductance (TCC) is one of the most important parameters in determining the temperature distribution in contacting structures. Thermal contact conductance between the contacting structures depends on the mechanical properties of underlying materials, thermo-physical properties of the interstitial fluid and surface condition of the structures coming in contact. During a postulated accident scenario of loss of coolant with coincident loss of emergency core cooling system in a tube type heavy water nuclear reactor, the pressure tube is expected to sag/balloon and come in contact with outer cooler calandria tube to dissipate away the heat generated to the moderator. The amount of heat thus transferred is a function of thermal contact conductance and the nature of contact between the two tubes. An experimental facility was designed, fabricated and commissioned to measure thermal contact conductance between pressure tube and calandria tube specimens. Experiments were conducted on disc shaped specimens under axial contact pressure in between mandrels. Experimental results of TCC and a linear correlation as a function of contact pressure have been reported in this paper.

  13. Canadian heavy water production

    International Nuclear Information System (INIS)

    The paper reviews Canadian experience in the production of heavy water, presents a long-term supply projection, relates this projection to the anticipated long-term electrical energy demand, and highlights principal areas for further improvement that form the bulk of our research and development program on heavy water processes

  14. Nuclear power plant life management processes: Guidelines and practices for heavy water reactors. Report prepared within the framework of the Technical Working Groups on Advanced Technologies for Heavy Water Reactors and on Life Management of Nuclear Power Plants

    International Nuclear Information System (INIS)

    The time is right to address nuclear power plant life management and ageing management issues in terms of processes and refurbishments for long term operation and license renewal aspects of heavy water reactors (HWRs) because some HWRs are close to the design life. In general, HWR nuclear power plant (NPP) owners would like to keep their NPPs in service as long as they can be operated safely and economically. This involves the consideration of a number of factors, such as the material condition of the plant, comparison with current safety standards, the socio-political climate and asset management/ business planning considerations. This TECDOC deals with organizational and managerial means to implement effective plan life management (PLiM) into existing plant in operating HWR NPPs. This TECDOC discusses the current trend of PLiM observed in NPPs to date and an overview of PLiM programmes and considerations. This includes key objectives of such programs, regulatory considerations, an overall integrated approach, organizational and technology infrastructure considerations, importance of effective plant data management and finally, human issues related to ageing and finally integration of PLiM with economic planning. Also general approach to HWR PLiM, including the key PLiM processes, life assessment for critical structures and components, conditions assessment of structures and components and obsolescence is mentioned. Technical aspects are described on component specific technology considerations for condition assessment, example of a proactive ageing management programme, and Ontario power generation experiences in appendices. Also country reports from Argentina, Canada, India, the Republic of Korea and Romania are attached in the annex to share practices and experiences to PLiM programme. This TECDOC is primarily addressed to both the management (decision makers) and technical staff (engineers and scientists) of NPP owners/operators and technical support

  15. 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

  16. Generic safety issues for nuclear power plants with pressurized heavy water reactors and measures for their resolution

    International Nuclear Information System (INIS)

    be used in reassessing the safety of individual operating plants. In 1998, the IAEA completed IAEA-TECDOC-1044 entitled Generic Safety Issues for Nuclear Power Plants with Light Water Reactors and Measures Taken for their Resolution and established the associated LWRGSIDB database (Computer Manual Series No. 13). The present compilation, which is based on broad international experience, is an extension of this work to cover pressurized heavy water reactors (PHWRs). As in the case of LWRs, it is one element in the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. It addresses generic safety issues identified in nuclear power plants using PHWRs. In most cases, the measures taken or planned to resolve these issues are also identified. The work on this report was initiated by the Senior Regulators of Countries Operating CANDU-Type Nuclear Power Plants at one of their annual meetings. It was carried out within the framework of the IAEA's programme on National Regulatory Infrastructure for Nuclear Installation Safety and serves to enhance regulatory effectiveness through the exchange of safety related information

  17. Controllability of depth dose distribution for neutron capture therapy at the Heavy Water Neutron Irradiation Facility of Kyoto University Research Reactor.

    Science.gov (United States)

    Sakurai, Yoshinori; Kobayashi, Tooru

    2002-10-01

    The updating construction of the Heavy Water Neutron Irradiation Facility of the Kyoto University Research Reactor has been performed from November 1995 to March 1996 mainly for the improvement in neutron capture therapy. On the performance, the neutron irradiation modes with the variable energy spectra from almost pure thermal to epi-thermal neutrons became available by the control of the heavy-water thickness in the spectrum shifter and by the open-and-close of the cadmium and boral thermal neutron filters. The depth distributions of thermal, epi-thermal and fast neutron fluxes were measured by activation method using gold and indium, and the depth distributions of gamma-ray absorbed dose rate were measured using thermo-luminescent dosimeter of beryllium oxide for the several irradiation modes. From these measured data, the controllability of the depth dose distribution using the spectrum shifter and the thermal neutron filters was confirmed.

  18. Integrated modular water reactor: IMR

    International Nuclear Information System (INIS)

    The Mitsubishi Heavy Industries, Ltd. Has investigated on a concept on small scale reactor with economical efficiency comparable with large scale one. Aims of development on the integrated modular water reactor (IMR) of a small scale reactor plant concept consist in large construction cost reduction through adoption of technique specific to the small scale reactor and integrated production of plural units and in establishment of high safety target without reality in a large scale reactor to realize reduction of operation and maintenance costs by this reduction to simplification of operation and maintenance. Its concrete developmental targets are to make an integrated reactor with vessel size actually producible and the largest output, to remove feasibility of coolant loss accident (LOCA), to remove an accident with feasibility related to fuel fracture, to remove feasibility of nuclear reactor coolant to leak out from a storage vessel, to secure safety of plant without necessity of human and physical assistances from other plants at all on an accident, to make numbers of operators per unit output equal to those of large scale reactor, and to make working amounts at maintenance per unit output equal to large scale reactor by simplification of apparatus practice of rotation on main apparatus such as SG, and so on. Here were described on design concept and plan to realization. (G.K.)

  19. Functional and performance evaluation of 28 bar hot shutdown passive valve (HSPV) at integral test loop (ITL) for advanced heavy water reactor (AHWR)

    International Nuclear Information System (INIS)

    During reactor shutdown in advanced heavy water reactor (AHWR), core decay heat is removed by eight isolation condensers (IC) submerged in gravity driven water pool. Passive valves are provided on the down stream of each isolation condenser. On increase in steam drum pressure beyond a set value, these passive valves start opening and establish steam flow by natural circulation between the four steam drums and corresponding isolation condensers under hot shutdown and therefore they are termed as Hot Shut Down Passive Valves (HSPVs). The HSPV is a self acting type valve requiring no external energy, i.e. neither air nor electric supply for actuation. This feature makes the valve functioning independent of external systems such as compressed air supply or electric power supply, thereby providing inherent safety feature in line with reactor design philosophy. The high pressure and high temperature HSPV s for nuclear reactor use, are non-standard valves and therefore not manufactured by the valve industry worldwide. In the process of design and development of a prototype valve for AHWR, a 28 bar HSPV was configured and successfully tested at Integral Test Loop (ITL) at Engineering Hall No.7. During ten continuous experiments spread over 14 days, the HSPV has proved its functional capabilities and its intended use in decay heat removal system. The in-situ pressure setting and calibration aspect of HSPV has also been successfully established during these experiments. This report gives an insight into the HSPV's functional behavior and role in reactor decay heat removal system. The report not only provides the quantitative measure of performance for 28 bar HSPV in terms of valve characteristics, pressure controllability, linearity and hysteresis but also sets qualitative indicators for prototype 80 bar HSPV, being developed for AHWR. (author)

  20. Fast microwave-assisted preparation of a low-cost and recyclable carboxyl modified lignocellulose-biomass jute fiber for enhanced heavy metal removal from water.

    Science.gov (United States)

    Du, Zhaolin; Zheng, Tong; Wang, Peng; Hao, Linlin; Wang, Yanxia

    2016-02-01

    A low-cost and recyclable biosorbent derived from jute fiber was developed for high efficient adsorption of Pb(II), Cd(II) and Cu(II) from water. The jute fiber was rapidly pretreated and grafted with metal binding groups (COOH) under microwave heating (MH). The adsorption behavior of carboxyl-modified jute fiber under MH treatment (CMJFMH) toward heavy metal ions followed Langmuir isotherm model (R(2)>0.99) with remarkably high adsorption capacity (157.21, 88.98 and 43.98mg/g for Pb(II), Cd(II) and Cu(II), respectively). Also, CMJFMH showed fast removal ability for heavy metals in a highly significant correlation with pseudo second-order kinetics model. Besides, CMJFMH can be easily regenerated with EDTA-2Na solution and reused up to at least four times with equivalent high adsorption capacity. Overall, cheap and abundant production, rapid and facile preparation, fast and efficient adsorption of heavy metals and high regeneration ability can make the CMJFMH a preferred biosorbent for heavy metal removal from water.

  1. Thermal hydraulic aspects of steam drum level control philosophy for the natural circulation based heavy water reactor

    International Nuclear Information System (INIS)

    From safety considerations advanced nuclear reactors rely more and more on passive systems such as natural circulation for primary heat removal. A natural circulation based water reactor is relatively larger in size so as to reduce flow losses and channel type for proper flow distribution. From the size of steam drum considerations it has to be multi loop but has a common inlet header. Normally the turbine follows the reactor. This paper addresses the thermal hydraulic aspects of the steam drum pressure and level control philosophy for a four drum, natural circulation based, channel type boiling water advanced reactor. Three philosophies may be followed for drum control viz. individual drum control, one control drum approach and an average of all the four drums. For drum pressure control, the steam flow to the turbine is be regulated. A single point pressure control is better than individual drum pressure control. This is discussed in the paper. But the control point has to be at a place down steam the point where all steam line from individual drum meet. This may lead to different pressure in all the four drums depending on the power produced in the respective loops. The difference in pressure cannot be removed even if the four drums are directly connected through pipes. Also the pressure control scheme with/without interconnection is discussed. For level, the control of individual drum may not be normally possible because of common inlet header. As the frictional pressure drops in the large diameter downcomers are small as compared to elevation pressure drops, the level in all the steam drum tend to equalize. Consequently a single representative drum level may be chosen as a control variable for controlling level in all the four drums. But in case, where all the four loops are producing different powers and single point pressure control is effective, the scheme may not work satisfactorily. the level in a drum may depend on the power produced in the loop

  2. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  3. Cost analysis and economic comparison for alternative fuel cycles in the heavy water cooled canadian reactor (CANDU)

    International Nuclear Information System (INIS)

    Three main options in a CANDU fuel cycle involve use of: (1) natural uranium (0.711 weight percent U-235) fuel, (2) slightly enriched uranium (1.2 weight percent U-235) fuel, and (3) recovered uranium (0.83 weight percent U-235) fuel from light water reactor spent fuel. ORIGEN-2 computer code was used to identify composition of the spent fuel for each option, including the standard LWR fuel (3.3 weight percent U-235). Uranium and plutonium credit calculations were performed using ORIGEN-2 output. WIMSD-5 computer code was used to determine maximum discharge burnup values for each case. For the 3 cycles selected (natural uranium, slightly enriched uranium, recovered uranium), levelized fuel cycle cost calculations are performed over the reactor lifetime of 40 years, using unit process costs obtained from literature. Components of the fuel cycle costs are U purchase, conversion, enrichment, fabrication, SF storage, SF disposal, and reprocessing where applicable. Cost parameters whose effects on the fuel cycle cost are to be investigated are escalation ratio, discount rate and SF storage time. Cost estimations were carried out using specially developed computer programs. Share of each cost component on the total cost was determined and sensitivity analysis was performed in order to show how a change in a main cost component affects the fuel cycle cost. The main objective of this study has been to find out the most economical option for CANDU fuel cycle by changing unit prices and cost parameters

  4. ARGOS PHWR 380. Argentine offer of a safer pressurized heavy-water reactor of 380 MW. '...a many-eyed guardian...' concerned about nuclear power plant safety

    International Nuclear Information System (INIS)

    Reactor vendors in most countries have had lean pickings for the past decade, and ordering seems unlikely to show much growth until the shock wave from the Chernobyl accident has died away. Paradoxically, however, at least one firm sees a niche in the market. ENACE - the Empresa Nuclear Argentina de Centrales Electricas, or Argentine Nuclear Power Plant Corporation - is stepping out into the market place with a newly-designed 380 MWe nuclear power plant. The plant is equipped with a pressurized heavy-water reactor of the pressure vessel type (PHWR). ENACE has adopted new boundary design conditions and has embodied a number of special features to assure safety and economy in operation. The major shareholder in ENACE is the Argentine National Atomic Energy Commission (CNEA). ENACE is the architect-engineer for the NPP projects of the Argentine nuclear programme. It has a licensing agreement with Siemens AG's Kraftwerk Union AG, which is its minor shareholder. Under this agreement, ENACE has the right to use the Siemens-KWU PHWR technology, which was originally developed for the MZFR reactor in the Federal Republic of Germany, as well as their know-how in pressurized (light-) water reactors (PWRs) design and construction. The CNEA also has agreements with Atomic Energy of Canada Ltd. for the transfer of technology related to CANDU-type PTHWRs. The CNEA and ENACE have acquired considerable practical experience from the construction and operation of the 367 MWe Atucha I PHWR and the 648 MWe Embalse PTHWR; ENACE is currently building Argentina's third nuclear power plant, Atucha II, a 745 MWe PHWR. (author)

  5. Double-side active TiO2-modified nanofiltration membranes in continuous flow photocatalytic reactors for effective water purification

    International Nuclear Information System (INIS)

    Highlights: ► A novel CVD reactor for the developments of double side active TiO2 membranes. ► Double side active TiO2 membranes efficiently photodegrade organic pollutants. ► A photocatalytic membrane purification device for continuous flow water treatment. - Abstract: A chemical vapour deposition (CVD) based innovative approach was applied with the purpose to develop composite TiO2 photocatalytic nanofiltration (NF) membranes. The method involved pyrolytic decomposition of titanium tetraisopropoxide (TTIP) vapor and formation of TiO2 nanoparticles through homogeneous gas phase reactions and aggregation of the produced intermediate species. The grown nanoparticles diffused and deposited on the surface of γ-alumina NF membrane tubes. The CVD reactor allowed for online monitoring of the carrier gas permeability during the treatment, providing a first insight on the pore efficiency and thickness of the formed photocatalytic layers. In addition, the thin TiO2 deposits were developed on both membrane sides without sacrificing the high yield rates. Important innovation was also introduced in what concerns the photocatalytic performance evaluation. The membrane efficiency to photo degrade typical water pollutants, was evaluated in a continuous flow water purification device, applying UV irradiation on both membrane sides. The developed composite NF membranes were highly efficient in the decomposition of methyl orange exhibiting low adsorption-fouling tendency and high water permeability.

  6. Evaluation of N,N-dihexyl octanamide as an alternative extractant for the reprocessing of Advanced Heavy Water Reactor spent fuel

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor (AHWR) is being developed in India with the specific aim of utilizing thorium for power generation. AHWR sent fuel adds new dimensions to reprocessing by the presence of Pu along with 233U and Th in the spent fuel. This invokes the integration of PUREX and THOREX processes in some combination employing tri-n-butyl phosphate (TBP) as an extractant. However, separation scientists have identified certain problems with the use of TBP as extractant viz. third-phase formation and low separation factor (SF) values of U(VI) and Pu(VI) over Th, and poor decontamination factor (DF) values of U and Pu with respect to fission products. These problems are of particular concern in thorium fuel cycle

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-30

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

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

    International Nuclear Information System (INIS)

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

  9. Removal of heavy metals from artificial metals contaminated water samples based on micelle-templated silica modified with pyoverdin I

    Institute of Scientific and Technical Information of China (English)

    PANADDA Tansupo; WORAKARN Chamonkolpradit; SAKSIT Chanthai; CHALERM Ruangviriyachai

    2009-01-01

    The micelle-templated silica (MTS) was firstly chemically modified with 3-glycidoxypropyl-trimethoxysilane (GPTMS) before immobilized with pyoverdin I. The characteristics of pyoverdin I-anchored onto the modified MTS were investigated using fluorescence, infrared spectra and scanning electron microscopy. The specific surface area of all materials was calculated by Bnmauer, Emmett and Teller (BET) method using nitrogen isotherm adsorption data. As the results, the surface area of commercial silica gel decreased from 609.2 to 405.4 m2/g, it indicated that the pyoverdin I could be immobilized onto the surface of silica solid support. This adsorbent was used for extraction of Fe(Ⅲ), Cu(Ⅱ), Zn(Ⅱ), and Pb(Ⅱ) in artificial metals contaminated water. Experimental conditions for effective adsorption of trace levels of metal ions were optimized with respect to different experimental parameters using batch procedure. The optimum pH value for the removal of metal ions simultaneously on this adsorbent was 4.0. Complete desorption of the adsorbed metal ions from the adsorbent was carded out using 0.25 mol/L of EDTA. The effcct of different cations and anions on the adsorption of these metals on adsorbent was studied and the results showed that the proposed adsorbent could be applied to the highly saline samples and the sample which contains some transition metals.

  10. Thorium utilization in heavy water moderated Accelerator Driven Systems

    International Nuclear Information System (INIS)

    Research on Accelerator Driven Systems (ADSs) is being carried out around the world primarily with the objective of waste transmutation. Presently, the volume of waste in India is small and therefore there is little incentive to develop ADS based waste transmutation technology immediately. With limited indigenous U availability and the presence of large Th deposits in the country, there is a clear incentive to develop Th related technologies. India also has vast experience in design, construction and operation of heavy water moderated reactors. Heavy water moderated reactors employing solid Th fuels can be self sustaining, but the discharge burnups are too low to be economical. A possible way to improve the performance such reactors is to use an external neutron source as is done in ADS. This paper discusses our studies on Th utilization in heavy water moderated ADSs. The study is carried out at the lattice level. The time averaged k-infinity of the Th bundle from zero burnup up to the discharge burnup is taken to be the same as the core (ensemble) averaged k-infinity. For the purpose of the analysis we have chosen standard PHWR and AHWR assemblies. Variation of the pitch and coolant (H2O/D2O) are studied. Both, the once through cycle and the recycling option are studied. In the latter case the study is carried out for various enrichments (% 233U in Th) of the recycled Th fuel bundles. The code DTF as modified for lattice and burnup calculations (BURNTRAN) was used for carrying out the study. The once through cycle represents the most attractive ADS concept (Th burner ADS) possible for Th utilization. It avoids reprocessing of Th spent fuel and in the ideal situation the use of any fissile material either initially or for sustaining itself. The gain in this system is however rather low requiring a high power accelerator and a substantial fraction of the power generated to be fed back to the accelerator. The self sustaining Th-U cycle in a heavy moderated ADS is a

  11. Novel modified pectin for heavy metal adsorption

    Institute of Scientific and Technical Information of China (English)

    Feng Ting Li; Hong Yang; Yan Zhao; Ran Xu

    2007-01-01

    Modified pectin cross-linked with adipic acid, was synthesized and used for heavy metal removal from wastewater. SEM and FrIR were used to investigate its structure and morphology. The modified pectin had a rough, porous phase covered with carboxy groups, resulting a high adsorption capacity. And at the room temperature, the saturated loading capacity for Pb2+, Cu2+ and Zn2+ reached 1.82 mmol/g, 1.794 mmol/g and 0.964 mmol/g, respectively. The results proved its potential application to remove of the heavy metal.

  12. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    International Nuclear Information System (INIS)

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs

  13. Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing 235U, 233U, and 232Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

    Science.gov (United States)

    Ioffe, B. L.; Kochurov, B. P.

    2012-02-01

    A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of 235U. It operates in the open-cycle mode involving 233U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

  14. Evaluation and analysis of critical crack length of irradiated pressure tubes from Indian pressurised heavy water reactors

    International Nuclear Information System (INIS)

    Results of fracture toughness KJic were computed from transverse tensile properties of reactor operated pressure tubes, and axial critical crack length values derived from KJic are presented. Similarly fracture resistance curves derived from tensile properties of reactor operated pressure tubes and axial critical crack length values computed therefrom are presented. Under normal operating condition of the reactors the pressure tubes experience temperatures ranging from 250 deg C to 300 deg C. In occurrence of contact between pressure tube and calandria tube, the contact region may not be expected to have a mean through wall temperature below 200 deg C. The axial critical crack length of three reactor operated pressure tubes, therefore were evaluated in the temperature range 200 deg C to 300 deg C. The significance of the magnitude of the evaluated critical crack length is discussed. (author)

  15. Double-side active TiO2-modified nanofiltration membranes in continuous flow photocatalytic reactors for effective water purification.

    Science.gov (United States)

    Romanos, G Em; Athanasekou, C P; Katsaros, F K; Kanellopoulos, N K; Dionysiou, D D; Likodimos, V; Falaras, P

    2012-04-15

    A chemical vapour deposition (CVD) based innovative approach was applied with the purpose to develop composite TiO(2) photocatalytic nanofiltration (NF) membranes. The method involved pyrolytic decomposition of titanium tetraisopropoxide (TTIP) vapor and formation of TiO(2) nanoparticles through homogeneous gas phase reactions and aggregation of the produced intermediate species. The grown nanoparticles diffused and deposited on the surface of γ-alumina NF membrane tubes. The CVD reactor allowed for online monitoring of the carrier gas permeability during the treatment, providing a first insight on the pore efficiency and thickness of the formed photocatalytic layers. In addition, the thin TiO(2) deposits were developed on both membrane sides without sacrificing the high yield rates. Important innovation was also introduced in what concerns the photocatalytic performance evaluation. The membrane efficiency to photo degrade typical water pollutants, was evaluated in a continuous flow water purification device, applying UV irradiation on both membrane sides. The developed composite NF membranes were highly efficient in the decomposition of methyl orange exhibiting low adsorption-fouling tendency and high water permeability. PMID:21999989

  16. Decontamination and dismantlement of the JANUS Reactor at Argonne National Laboratory-East. Project final report

    Energy Technology Data Exchange (ETDEWEB)

    Fellhauer, C.R.; Clark, F.R. [Argonne National Lab., IL (United States). Technology Development Div.; Garlock, G.A. [MOTA Corp., Cayce, SC (United States)

    1997-10-01

    The decontamination and dismantlement of the JANUS Reactor at Argonne National Laboratory-East (ANL-E) was completed in October 1997. Descriptions and evaluations of the activities performed and analyses of the results obtained during the JANUS D and D Project are provided in this Final Report. The following information is included: objective of the JANUS D and D Project; history of the JANUS Reactor facility; description of the ANL-E site and the JANUS Reactor facility; overview of the D and D activities performed; description of the project planning and engineering; description of the D and D operations; summary of the final status of the JANUS Reactor facility based upon the final survey results; description of the health and safety aspects of the project, including personnel exposure and OSHA reporting; summary of the waste minimization techniques utilized and total waste generated by the project; and summary of the final cost and schedule for the JANUS D and D Project.

  17. Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the 233U isotope in the VVER reactors using thorium and heavy water

    Science.gov (United States)

    Marshalkin, V. E.; Povyshev, V. M.

    2015-12-01

    A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium-uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D2O, H2O) is proposed. The method is characterized by efficient breeding of the 233U isotope and safe reactor operation and is comparatively simple to implement.

  18. Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the 233U isotope in the VVER reactors using thorium and heavy water

    International Nuclear Information System (INIS)

    A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium–uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D2O, H2O) is proposed. The method is characterized by efficient breeding of the 233U isotope and safe reactor operation and is comparatively simple to implement

  19. Status of the Argonne heavy-ion-fusion low-beta linac

    Energy Technology Data Exchange (ETDEWEB)

    Watson, J.M.; Bogaty, J.M.; Moretti, A.; Sacks, R.A.; Sesol, N.Q.; Wright, A.J.

    1981-01-01

    The primary goal of the experimental program in heavy-ion fusion (HIF) at Argonne National Laboratory (ANL) during the next few years is to demonstrate many of the requirements of a RF linac driver for inertial-fusion power plants. So far, most of the construction effort has been applied to the front end. The ANL program has developed a high-intensity xenon source, a 1.5-MV preaccelerator, and the initial cavities of the low-beta linac. The design, initial tests, and status of the low-beta linac are described.

  20. Status of the Argonne heavy ion fusion low-beta linac

    Energy Technology Data Exchange (ETDEWEB)

    Watson, J.M.; Bogaty, J.M.; Moretti, A.; Sacks, R.A.; Sesol, N.Q.; Wright, A.J.

    1981-06-01

    The primary goal of the experimental program in heavy ion fusion (HIF) at Argonne National Laboratory (ANL) during the next few years is to demonstrate many of the requirements of a RF linac driver for inertial fusion power plants. So far, most of the construction effort has been applied to the front end. The ANL program has developed a high intensity xenon source, a 1.5 MV preaccelerator, and the initial cavities of the low-beta linac. The design, initial tests and status of the low-beta linac are described. 8 refs.

  1. Development of poison injection code-COPJET for high pressure liquid poison injection in pressure tube type heavy water reactor

    International Nuclear Information System (INIS)

    Shut Down System-2 (SDS-2) in advanced vertical pressure tube type reactor, provides rapid reactor shutdown by high pressure injection of a neutron absorbing liquid called poison, into the moderator in the calandria. Poison inside the calandria is distributed by poison jets issued from holes provided in the injection tubes. Effectiveness of the system depends on the rate and spread of the poison in the moderator. In this study, a transient one-dimensional (1-D) hydraulic code, COPJET is developed, to predict the performance of system by predicting poison jet length with time. Validation of the COPJET is done with the data available in literature. Thereafter, it is applied for poison jet length prediction of advanced vertical pressure type reactor. (author)

  2. Plasma coating used to evaluate resistance against flow accelerated corrosion on carbon steel feeder pipe material for pressurized heavy water reactor

    International Nuclear Information System (INIS)

    A collaborative study on plasma nitriding was initiated by Water and Steam Chemistry Division, BARC Facilities, Kalpakkam, with FCIPT, a division of Institute of Plasma Research. In order to control the influence of Flow Accelerated Corrosion (FAC) on feeder pipe of PHWR reactor, coating by plasma nitriding process was carried out inside the pipe as a remedy.This is one of the methods to control the wall thickness reduction of carbon steel feeder pipe and the influence of FAC in PHWR (Pressurized heavy water reactor). Specimen of 15 mm NB Sch 80 straight pipe length of 100 mm pipe module section of low carbon steel ASTM 106 Gr. B were plasma nitrided at FCIPT, IPR for optimization of the process parameters. The wall thickness of the sample was measured axially and circumferentially by Ultrasonic thickness gauge with specific marking with templates before carrying out plasma nitriding process. During plasma nitriding the temperature was maintained at 520 °C for 24 hours. The samples after coating were checked for thickness variation by Raman spectroscopy as well as microscopy, and it was found that the coating was uniform and coating consisted of iron nitrides only. For functional test, to check the corrosion resistance, a specimen holder was designed and fabricated for the treated specimen such that it can withstand a velocity of 7 m/s. The holder was mounted in SIM loop outlet of heater. The SIM loop was maintained at 120 °C and 7 m/s for about 30 days with less than 20 ppb dissolved oxygen condition. Preliminary experiments on plasma nitriding have been carried out and checked in SIM loop in order to check the resistance to FAC under neutral pH condition. (author)

  3. Efficiency of Algae Combinations in heavy metal removal from waste-waters using photo-bio-reactor

    OpenAIRE

    Bello, Adedayo

    2015-01-01

    The aim of this thesis was to compare the efficiency of different algal combinations in heavy metals removal from wastewater using algae-based photo-bioreactors. Twelve different strains of algae were divided into four groups and were introduced into twenty-four photo-bioreactor bottles: twelve contained wastewaters only while the other twelve contained wastewaters con-taminated with 90 mg of heavy metal. Parameters such as temperature, pH, light and conductivi-ty, which are believed to affec...

  4. Specialists' meeting on advanced controls for fast reactors, Argonne, Illinois, USA June 20-22, 1989

    International Nuclear Information System (INIS)

    The Specialists' Meeting on ''Advanced Controls for Fast Reactors'' was held in Argonne, Illinois, USA, from June 20 to 22, 1989. The meeting was sponsored by the International Atomic Energy Agency on the recommendation of the IAEA International Working Group on Fast Reactors and was hosted by Argonne National Laboratory and the US Department of Energy. It was attended by 20 participants and observers from Argentina, France, Germany, Japan, India, the USSR, the United Kingdom, the United States of America, and the IAEA. The purpose of the meeting was to provide an opportunity for participating countries to review and discuss their views on design and technology for advanced control in fast reactors. During the meeting papers were presented by the participants on behalf of their countries and organizations. Presentations were followed by open discussions on the subjects covered by the papers and summaries of the discussions were drafted. After the formal sessions were completed, a final discussion session was held and summaries, general conclusions and recommendations were approved by consensus. A separate abstract was prepared for each of the 22 papers presented at this meeting. Refs, figs, tabs, diagrams and photos

  5. Environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with ∼ 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289 degrees C

  6. Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water

    Energy Technology Data Exchange (ETDEWEB)

    Marshalkin, V. E., E-mail: marshalkin@vniief.ru; Povyshev, V. M. [Russian Federal Nuclear Center All-Russian Research Institute of Experimental Physics (Russian Federation)

    2015-12-15

    A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium–uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D{sub 2}O, H{sub 2}O) is proposed. The method is characterized by efficient breeding of the {sup 233}U isotope and safe reactor operation and is comparatively simple to implement.

  7. Addendum to a proposal for ATLAS: a precision heavy-ion accelerator at Argonne National Laboratory

    International Nuclear Information System (INIS)

    This revised proposal for the construction of the Argonne Tandem-Linac Accelerator System (ATLAS) is in all essentials the same as the proposal originally presented to NUSAC in March 1978. The only differences worth mentioning are the plan to expand the experimental area somewhat more than was originally proposed and an increased cost, brought about principally by inflation. The outline presented is the same as in the original document, reproduced for the convenience of the reader. The objective of the proposed Argonne Tandem-Linac Accelerator System (ATLAS) is to provide precision beams of heavy ions for nuclear physics research in the region of projectile energies comparable to nuclear binding energies (5 to 25 MeV/A). By using the demonstrated potential of superconducting rf technology, beams of exceptional quality and flexibility can be obtained. The proposed system is designed to provide beams with tandem-like energy resolution and ease of energy variation, and the energy range is comparable to that of a approx. 50 MV tandem. In addition, the beam will be bunched into very short (approx. 50 psec) pulses, permitting fast-timing measurements that can open up major new experimental approaches

  8. Environmentally assisted cracking in Light Water Reactors

    International Nuclear Information System (INIS)

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289 degrees C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy

  9. Improvement of Severe Accident Analysis Computer Code and Development of Accident Management Guidance for Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Kim, Ko Ryu; Kim, Dong Ha; Kim, See Darl; Song, Yong Mann; Choi, Young; Jin, Young Ho

    2005-03-15

    The objective of the project is to develop a generic severe accident management guidance(SAMG) applicable to Korean PHWR and the objective of this 3 year continued phase is to construct a base of the generic SAMG. Another objective is to improve a domestic computer code, ISAAC (Integrated Severe Accident Analysis code for CANDU), which still has many deficiencies to be improved in order to apply for the SAMG development. The scope and contents performed in this Phase-2 are as follows: The characteristics of major design and operation for the domestic Wolsong NPP are analyzed from the severe accident aspects. On the basis, preliminary strategies for SAM of PHWR are selected. The information needed for SAM and the methods to get that information are analyzed. Both the individual strategies applicable for accident mitigation under PHWR severe accident conditions and the technical background for those strategies are developed. A new version of ISAAC 2.0 has been developed after analyzing and modifying the existing models of ISAAC 1.0. The general SAMG applicable for PHWRs confirms severe accident management techniques for emergencies, provides the base technique to develop the plant specific SAMG by utility company and finally contributes to the public safety enhancement as a NPP safety assuring step. The ISAAC code will be used inevitably for the PSA, living PSA, severe accident analysis, SAM program development and operator training in PHWR.

  10. Status of advanced technology and design for water cooled reactors: Light water reactors

    International Nuclear Information System (INIS)

    Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of 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 different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

  11. Optimisation of the hot conditioning of carbon steel surfaces of primary heat transport system of Pressurized Heavy Water Reactors using electrochemical impedance spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Kiran Kumar, M., E-mail: mkiran@barc.gov.i [Materials Science Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Gaonkar, Krishna; Ghosh, Swati; Kain, Vivekanand [Materials Science Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Bojinov, Martin [Department of Physical Chemistry, University of Chemical Technology and Metallurgy, Kl. Ohridski Blvd. 8, 1756 Sofia (Bulgaria); Saario, Timo [VTT Materials and Building, VTT Technical Research Centre of Finland, P.O. Box 1000, Kemistintie 3, FIN-02044 VTT, Espoo (Finland)

    2010-06-15

    Hot conditioning operation of the primary heat transport system is an important step prior to the commissioning of Pressurized Heavy Water Reactors. One of the major objectives of the operation is to develop a stable and protective magnetite layer on the inner surfaces of carbon steel piping. The correlation between stable magnetite film growth on carbon steel surfaces and the period of exposure to hot conditioning environment is generally established by a combination of weight change measurements and microscopic/morphological observations of the specimens periodically removed during the operation. In the present study, electrochemical impedance spectroscopy (EIS) at room temperature is demonstrated as an alternate, quantitative technique to arrive at an optimal duration of the exposure period. Specimens of carbon steel were exposed for 24, 35 and 48 h during hot conditioning of primary heat transport system of two Indian PHWRs. The composition and morphology of oxide films grown during exposure was characterized by X-ray diffraction and optical microscopy. Further, ex situ electrochemical impedance spectra of magnetite films formed after each exposure were measured, in 1 ppm Li{sup +} electrolyte at room temperature as a function of potential in a range of -0.8 to +0.3 V{sub SCE}. The defect density of the magnetite films formed after each exposure was estimated by Mott-Schottky analysis of capacitances extracted from the impedance spectra. Further the ionic resistance of the oxide was also extracted from the impedance spectra. Defect density was observed to decrease with increase in exposure time and to saturate after 35 h, indicating stabilisation of the barrier layer part of the magnetite film. The values of the ionic transport resistance start to increase after 35-40 h of exposure. The quantitative ability of EIS technique to assess the film quality demonstrates that it can be used as a supplementary tool to the thickness and morphological characterizations of

  12. Best practices in management of heavy water and tritium

    International Nuclear Information System (INIS)

    The heavy water inventory of a typical HWR constitutes about 12% of the capital cost of the HWR. The typical tritium production in a single unit HWR is about 2 x 106 Ci/y.1 Heavy water and tritium control are important aspects of HWR operation, and this involves people, procedures, equipment and heavy water and tritium separation systems. Station personnel are trained to understand the importance of heavy water management and the economics and environmental impact of tritiated heavy water losses. The tritium and heavy water losses from a HWR are both airborne and waterborne in nature. Tritium is of particular concern in the HWR industry given the nature of heavy water reactors to build up high levels of tritium over time. Recent increased interest from regulators and the public has led more HWR utilities to pay increasing attention to occupational safety and environmental emissions of tritium at their power stations. As competing reactor technologies improve, a simple and economic means for tritium removal from heavy water in HWRs is essential for the long- term attractiveness of HWR technology. Tritium safety, occupational and environmental issues are of central importance in HWR licensing and operation. Building upon GE's extensive operational experience in tritium management in HWR reactors and its own tritium handling facility, GE2 has developed a large-scale diffusion-based isotope separation process as an alternative to conventional cryogenic distillation. Having a tritium inventory an order of magnitude lower than conventional cryogenic distillation, this process is very attractive for heavy water detritiation, applicable to single and multi-unit HWR and research reactors. Additionally, the new process has significant benefits to an operating HWR utility such as reducing environmental emissions and significantly lowering reactor vault tritium MPC(a) levels to a point where station capacity factors can be improved by shorter outages - representing best

  13. Advanced liquid metal reactor development at Argonne National Laboratory during the 1980s

    International Nuclear Information System (INIS)

    Argonne National Laboratory's (ANL'S) effort to pursue the exploitation of liquid metal cooled reactor (LMR) characteristics has given rise to the Integral Fast Reactor (IFR) concept, and has produced substantial technical advancement in concept implementation which includes demonstration of high burnup capability of metallic fuel, demonstration of injection casting fabrication, integral demonstration of passive safety response, and technical feasibility of pyroprocessing. The first half decade of the 90's will host demonstration of the IFR closed fuel cycle technology at the prototype scale. The EBR-II reactor will be fueled with ternary alloy fuel in HT-9 cladding and ducts, and pyroprocessing and injection casting refabrication of EBR-II fuel will be conducted using near-commercial sized equipment at the Fuel cycle Facility (FCF) which is co-located adjacent to EBR-II. Demonstration will start in 1992. The demonstration of passive safety response achievable with the IFR design concept, (already done in EBR-II in 1986) will be repeated in the mid 90's using the IFR prototype recycle fuel from the FCF. The demonstration of scrubbing of the reprocessing fission product waste stream, with recycle of the transuranics to the reactor for consumption, will also occur in the mid 90's. 30 refs

  14. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    The principles of nuclear, thermal and hydraulic performance calculations for water cooled reactors are discussed. The principles are illustrated by describing their implementation in the UKAEA PATRIARCH scheme of computer codes. This material was originally delivered as a course of lectures at the Technical University of Helsinki in Summer of 1969.

  15. Effective Removal of Heavy Metals from Wastewater Using Modified Clay.

    Science.gov (United States)

    Song, Mun-Seon; Vijayarangamuthu, K; Han, EunJi; Jeon, Ki-Joon

    2016-05-01

    We report an economical and eco-friendly way to remove the heavy metal pollutant using modified clay. The modification of clay was done by calcining the natural clay from Kyushu region in Japan. Further, the removal efficiency for various pH and contact time was evaluated. The morphology of the clays was studied using the scanning electron microscopy (SEM). The structural and chemical analyses of modified clay were done by using X-ray diffraction (XRD), Raman spectroscopy, and Energy dispersion analysis (EDAX) to understand the properties related to the removal of heavy metal pollutant. Further, we studied the absorption efficiency of clay for various pH and contacting time using Ni polluted water. The modified clays show better removal efficiency for all pH with different saturation time. The adsorption follows pseudo-second order kinetics and the adsorption capacity of modified clay is 1.5 times larger than that of natural clay. The increase in the adsorption efficiency of modified clay was correlated to the increase in hematite phase along with increase in surface area due to surface morphological changes.

  16. Environmentally assisted cracking in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E. [and others

    1996-07-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  17. Removal of gadolinium nitrate from heavy water

    Energy Technology Data Exchange (ETDEWEB)

    Wilde, E.W.

    2000-03-22

    Work was conducted to develop a cost-effective process to purify 181 55-gallon drums containing spent heavy water moderator (D2O) contaminated with high concentrations of gadolinium nitrate, a chemical used as a neutron poison during former nuclear reactor operations at the Savannah River Site (SRS). These drums also contain low level radioactive contamination, including tritium, which complicates treatment options. Presently, the drums of degraded moderator are being stored on site. It was suggested that a process utilizing biological mechanisms could potentially lower the total cost of heavy water purification by allowing the use of smaller equipment with less product loss and a reduction in the quantity of secondary waste materials produced by the current baseline process (ion exchange).

  18. Heavy water detritiation to support CANDU station maintenance activities

    International Nuclear Information System (INIS)

    Heavy water and tritium control are important aspects of CANDU operation and have a major impact on station maintenance activities. Station personnel are trained to understand the importance of heavy water management and the economics and environmental impact of tritiated heavy water losses. This paper discusses new GE technology that can now make a major improvement in CANDU maintenance activities through significant reductions in station tritium levels. Tritium is of particular concern in the CANDU industry given the nature of heavy water reactors to build up high levels of tritium over time. High tritium levels in the reactor vault significantly slow down maintenance activities in the reactor vault due to the requirement for personnel protective equipment, including breathing apparatus and cumbersome plastic air suits. The difficulties increase as reactors age and tritium levels increase. Building upon GE's extensive operational experience in tritium management in CANDU reactors and its own tritium handling facility, GE. has developed a new large-scale diffusion based isotope separation process as an alternative to conventional cryogenic distillation. Having a tritium inventory an order of magnitude lower than conventional cryogenic distillation, this process is very attractive for heavy water detritiation, and applicable to single and multi-unit CANDU stations. This new process can now provide a step change reduction in CANDU heavy water tritium levels resulting in reduced environmental emissions and lowering reactor vault tritium MPC(a) levels. Reactor vault tritium can be reduced sufficiently for maintenance activities to be done without plastic suits, leading to shorter outages, improved station capacity factors, and improved station economics. (author)

  19. Pressurized Heavy Water Reactor Fuel: Integrity, Performance and Advanced Concepts. Proceedings of the Technical Meetings held in Bucharest, 24-27 September 2012, and in Mumbai, 8-11 May 2013

    International Nuclear Information System (INIS)

    Seven Member States have operating pressurized heavy water reactors (PHWRs), and some of them are also planning new reactors of this type. The current type of PHWR uses natural uranium as the fuel and has an average burnup of 7000 MWd/t (megawatt days per metric tonne). To make these reactors economically competitive with other reactor types, the discharge burnup of PHWR fuel will need to be increased without affecting the integrity of the fuel pin and bundle. A significant increase in the discharge burnup of fuel is possible with the use of advanced fuel cycles in PHWRs. The advanced fuels can be slightly enriched uranium, reprocessed uranium from light water reactors, mixed oxide or thorium based fuels. At the same time, substantial savings in natural uranium resources can also be achieved through the possible extension of the discharge burnup of advanced fuels used in PHWRs without changing reactor hardware. Following the recommendation of the Technical Working Group on Fuel Performance and Technology, two technical meetings were held: Technical Meeting on Fuel Integrity during Normal Operation and Accident Conditions in PHWRs, 24–27 September 2012, Bucharest, Romania; and Technical Meeting on Advanced Fuel Cycles in PHWRs, 8–11 April 2013, Mumbai, India. Their objective was to update information on the performance of PHWR fuels, the status and trends in the use of advanced fuels in PHWRs and the technical readiness for the deployment of such fuel cycles in these types of reactor. This publication contains the proceedings of the two technical meetings, including a record of the discussions held during the various technical sessions

  20. Light-Water-Reactor safety research program. Quarterly progress report, January--March 1977

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-06-01

    The report summarizes the Argonne National Laboratory work performed during January, February, and March 1977 on water-reactor-safety problems. The following research and development areas are covered: (1) loss-of-coolant accident research: heat transfer and fluid dynamics; (2) transient fuel response and fission-product release program; (3) mechanical properties of zircaloy containing oxygen; and (4) steam-explosion studies.

  1. Heavy ion de-acceleration with the Argonne Tandem-Linac Accelerator

    International Nuclear Information System (INIS)

    The Argonne Tandem-Linac Accelerator system has been used to produce beams of 0.375 MeV/A 16O 8+ and 0.386 MeV/A 28Si 13+ and 28Si 14+ as a test of using the superconducting linac de-acceleration mode to provide highly stripped high charge state heavy-ion beams for use in atomic physics experimental programs. Such beams have been developed in the past at installations containing dual tandem electrostatic accelerators and the U. of Heidelberg tandem-linac facility. The beams in the tests reported in this communication were transmitted through the linac with an efficiency of 30 to 50% and can be delivered to a target location with a transmission efficiency of approximately 7%. These tests required the use of only 50 to 75% of the present linac. Energies down to 0.135 MeV/A should be possible using the entire linac but these lower energies will be accompanied by significant additional losses in transmission efficiency due to longitudinal and transverse emittance growth

  2. Feasibility and deployment strategy of water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    The author have studied water cooled thorium breeder reactor based on matured pressurized water reactor (PWR) plant technology for several years. Through these studies it is concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array with heavy water coolant in the primary loop by replacing original light water is appropriate for achieving sufficient breeding performance as sustainable fission system and high enough burn-up as an economical power plant. The heavy water cooled thorium reactor is feasible to be introduced by using Pu recovered from spent fuel of LWR, keeping continuity with current LWR infrastructure. This thorium reactor can be operated as sustainable energy supplier and also MA transmuter to realize future society with less long-lived nuclear waste

  3. Direction of Heavy Water Projects

    International Nuclear Information System (INIS)

    Summary of the activities performed by the Heavy Water Projects Direction of the Argentine Atomic Energy Commission from 1950 to 1983. It covers: historical data; industrial plant (based on ammonia-hydrogen isotopic exchange); experimental plant (utilizing hydrogen sulfides-water process); Module-80 plant (2-3 tons per year experimental plant with national technology) and other related tasks on research and development (E.A.C.)

  4. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  5. Study relating to the physico-chemical behaviour of heavy water in nuclear reactors; Etudes relatives au comportement physico-chimique de l'eau lourde dans les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Chenouard, J.; Dirian, G.; Roth, E.; Vignet, P.; Platzer, R. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1959-07-01

    Chemical and isotope pollution, and radiolytic decomposition are the two most important ways in which heavy water becomes degraded in nuclear reactors. Chemical pollution has led to the creation of ion exchange purification loops specially designed for reactors: the report contains a description in detail of the application of this purification method in CEA research reactors, including the analysis required, results obtained, and their interpretation. The intelligence obtained on radiolytic decomposition with the same facilities is also discussed, as well as the recombination apparatus and control equipment utilized. Finally, investigation to date in the CEA on recombination circuits for power reactors is also discussed. (author) [French] Parmi les degradations subies par l'eau lourde dans les reacteurs nucleaires, les deux plus importantes sont la pollution chimique et isotopique et la decomposition radiolytique. La pollution chimique a conduit a mettre au point pour le cas particulier des reacteurs, des circuits d'epuration par echange d'ions. On decrit ici en detail la mise en oeuvre de cette methode dans les reacteurs de recherche du CEA; les controles qu'elle necessite, les resultats obtenus et leur interpretation. En ce qui concerne la dissociation radiolytique de l'eau, les renseignements obtenus sur ces memes reacteurs sont communiques, ainsi que les details des dispositifs de recombinaison et des moyens de controle. Enfin, on fait le point des etudes poursuivies au CEA sur ces memes problemes de recombinaison dans le cas des reacteurs de puissance. (auteur)

  6. Heavy water. A production alternative for Venezuela

    International Nuclear Information System (INIS)

    A survey of heavy water production methods is made. Main facts about isotopic and distillation methods, reforming and coupling to a Hydrogen distillation plant are presented. A feasibility study on heavy water production in Venezuela is suggested

  7. Heavy water at Trail, British Columbia

    International Nuclear Information System (INIS)

    Today Canada stands on the threshold of a nuclear renaissance, based on the CANDU reactor family, which depends on heavy water as a moderator and for cooling. Canada has a long history with heavy water, with commercial interests beginning in 1934, a mere two years after its discovery. At one time Canada was the world's largest producer of heavy water. The Second World War stimulated interest in this rather rare substance, such that the worlds largest supply (185 kg) ended up in Canada in 1942 to support nuclear research work at the Montreal Laboratories of the National Research Council. A year later commercial production began at Trail, British Columbia, to support work that later became known as the P-9 project, associated with the Manhattan Project. The Trail plant produced heavy water from 1943 until 1956, when it was shut down. During the war years the project was so secret that Lesslie Thomson, Special Liaison Officer reporting on nuclear matters to C.D. Howe, Minister of Munitions and Supply, was discouraged from visiting Trail operations. Thomson never did visit the Trail facility during the war. In 2005 the remaining large, tall concrete exchange tower was demolished at a cost of about $2.4 million, about the same as it cost to construct the facility about 60 years ago. Thus no physical evidence remains of this historic facility and another important artifact from Canada's nuclear history has disappeared forever. It is planned to place a plaque at the site at some point in the future. (author)

  8. Discovery of Interstellar Heavy Water

    OpenAIRE

    Butner, H. M.; Charnley, S. B.; Ceccarelli, C.; Rodgers, S.D.; Pardo Carrión, Juan Ramón; Parise, B.; Cernicharo, José; Davis, G. R.

    2007-01-01

    We report the discovery of doubly deuterated water (D2O, heavy water) in the interstellar medium. Using the James Clerk Maxwell Telescope and the Caltech Submillimeter Observatory 10 m telescope, we detected the 1_10–1_01 transition of para-D2O at 316.7998 GHz in both absorption and emission toward the protostellar binary system IRAS 16293-2422. Assuming that the D2O exists primarily in the warm regions where water ices have been evaporated (i.e., in a "hot corino" environment), we determi...

  9. Decontamination and decommissioning of the Argonne Thermal Source Reactor at Argonne National Laboratory - East project final report.

    Energy Technology Data Exchange (ETDEWEB)

    Fellhauer, C.; Garlock, G.; Mathiesen, J.

    1998-12-02

    The ATSR D&D Project was directed toward the following goals: (1) Removal of radioactive and hazardous materials associated with the ATSR Reactor facility; (2) Decontamination of the ATSR Reactor facility to unrestricted use levels; and (3)Documentation of all project activities affecting quality (i.e., waste packaging, instrument calibration, audit results, and personnel exposure). These goals had been set in order to eliminate the radiological and hazardous safety concerns inherent in the ATSR Reactor facility and to allow, upon completion of the project, unescorted and unmonitored access to the area. The reactor aluminum, reactor lead, graphite piles in room E-111, and the contaminated concrete in room E-102 were the primary areas of concern. NES, Incorporated (Danbury, CT) characterized the ATSR Reactor facility from January to March 1998. The characterization identified a total of thirteen radionuclides, with a total activity of 64.84 mCi (2.4 GBq). The primary radionuclides of concern were Co{sup 60}, Eu{sup 152}, Cs{sup 137}, and U{sup 238}. No additional radionuclides were identified during the D&D of the facility. The highest dose rates observed during the project were associated with the reactor tank and shield tank. Contact radiation levels of 30 mrem/hr (0.3 mSv/hr) were measured on reactor internals during dismantlement of the reactor. A level of 3 mrem/hr (0.03 mSv/hr) was observed in a small area (hot spot) in room E-102. DOE Order 5480.2A establishes the maximum whole body exposure for occupational workers at 5 rem/yr (50 mSv/yr); the administrative limit at ANL-E is 1 rem/yr (10 mSv/yr).

  10. Light water reactor safety research

    International Nuclear Information System (INIS)

    As the technology of light water reactors (LWR) was being commercialized, the German Federal Government funded the reactor safety research program, which was conducted by national research centers, universities, and industry, and which led to the establishment, in early 1972, of the Nuclear Safety Project in Karlsruhe. In the seventies, the PNS project mainly studied the loss-of-coolant accident. Numerous experiments were run and computer codes developed for this purpose. In the eighties, the Karlsruhe Nuclear Research Center contributed to the German Risk Study, investigating especially core meltdown accidents under the impact of the events at Three Mile Island-2 and Chernobyl-4. Safety research in the nineties is concentrated on the requirements of future reactor generations, such as the European Pressurized Water Reactor (EPR) or potential approaches which, at the present time, are discernible only as tentative theoretical designs. (orig.)

  11. Heavy water: a distinctive and essential component of CANDU

    International Nuclear Information System (INIS)

    The exceptional properties of heavy water as a neutron moderator provide one of the distinctive features of CANDU reactors. Although most of the chemical and physical properties of deuterium and protium (mass 1 hydrogen) are appreciably different, the low terrestrial abundance of deuterium makes the separation of heavy water a relatively costly process, and so of considerable importance to the CANDU system. World heavy-water supplies are currently provided by the Girdler-Sulphide process or processes based on ammonia-hydrogen exchange. Due to cost and hazard considerations, new processes will be required for the production of heavy water in and beyond the next decade. Through AECL's development and refinement of wetproofed catalysts for the exchange of hydrogen isotopes between water and hydrogen, a family of new processes is expected to be deployed. Two monothermal processes, CECE (Combined Electrolysis and Catalytic Exchange, using water-to-hydrogen conversion by electrolysis) and CIRCE (Combined Industrially Reformed hydrogen and Catalytic Exchange, based on steam reforming of hydrocarbons), are furthest advanced. Besides its use for heavy-water production, the CECE process is a highly effective technology for heavy-water upgrading and for tritium separation from heavy (or light) water. (author). 10 refs., 1 tab., 7 figs

  12. Water shielding nuclear reactor container

    International Nuclear Information System (INIS)

    The reactor container of the present invention contains a reactor pressure vessel, and has double steel plate walls endurable to elevated inner pressure and keeping airtightness, and shielding water is filled inside from a water injection port. It is endurable to a great inner pressure satisfactorily and keep airtightness by the two spaced relatively thin steel plates. It exhibits radiation shielding effect by filling water substantially the same as that of a conventional reactor container made of iron reinforced concretes. Then, it is no more necessary to use concretes for the construction of the reactor container, which shortens the term of the construction, and saves the construction cost. In addition, a cooling effect for the reactor container is provided. Syphons are disposed contiguously to a water injection port and the top end of the syphon is immersed in an equipment temporarily storage pool, and further, pipelines are connected to the double steel plate walls or the syphons for supplying shielding water to enhance the cooling effect. (N.H.)

  13. Researches on a reactor core in heavy ion inertial fusion

    CERN Document Server

    Kondo, S; Iinuma, T; Kubo, K; Kato, H; Kawata, S; Ogoyski, A I

    2016-01-01

    In this paper a study on a fusion reactor core is presented in heavy ion inertial fusion (HIF), including the heavy ion beam (HIB) transport in a fusion reactor, a HIB interaction with a background gas, reactor cavity gas dynamics, the reactor gas backflow to the beam lines, and a HIB fusion reactor design. The HIB has remarkable preferable features to release the fusion energy in inertial fusion: in particle accelerators HIBs are generated with a high driver efficiency of ~30-40%, and the HIB ions deposit their energy inside of materials. Therefore, a requirement for the fusion target energy gain is relatively low, that would be ~50 to operate a HIF fusion reactor with a standard energy output of 1GW of electricity. In a fusion reactor the HIB charge neutralization is needed for a ballistic HIB transport. Multiple mechanical shutters would be installed at each HIB port at the reactor wall to stop the blast waves and the chamber gas backflow, so that the accelerator final elements would be protected from the ...

  14. Modified wetted-wall inertial fusion reactor concept

    International Nuclear Information System (INIS)

    Limitations on reactor pulse repetition rate and uncertainties with respect to assurance of first wall protection in LASL wetted-wall inertial fusion reactor concepts, in which restoration of cavity conditions to those required for acceptable driver energy pulse transmission following pellet microexplosion is accomplished by exhaust of ablated liquid metal through nozzles and protective films are formed by forcing liquid metal through porous first walls, can be circumvented through alternative methods of cavity clearing and protective film formation. Exploratory analyses indicate that our modified wetted-wall concept, in which protective liquid metal films are injected directly onto cavity walls through slit nozzles to ensure first wall protection and are held there by centrifugal forces and cavity clearing occurs by condensation of vapor on film liquid not ablated as a result of pellet x ray and debris ion energy deposition, can be operated at substantially higher repetition rates. The new mode of operation appears to be attractive for heavy ion fusion, for which constraints on cavity design options may be more severe, as well as laser fusion. Numerical results of the exploratory analyses, plus discussion of aspects of the new concept requiring further work, are presented

  15. Double-side active TiO{sub 2}-modified nanofiltration membranes in continuous flow photocatalytic reactors for effective water purification

    Energy Technology Data Exchange (ETDEWEB)

    Romanos, G.Em., E-mail: groman@chem.demokritos.gr [Institute of Physical Chemistry, NCSR Demokritos, 153 10 Agia Paraskevi Attikis, Athens (Greece); Athanasekou, C.P.; Katsaros, F.K.; Kanellopoulos, N.K. [Institute of Physical Chemistry, NCSR Demokritos, 153 10 Agia Paraskevi Attikis, Athens (Greece); Dionysiou, D.D. [Department of Civil and Environmental Engineering, University of Cincinnati, Cincinnati, OH 45221-0071 (United States); Likodimos, V.; Falaras, P. [Institute of Physical Chemistry, NCSR Demokritos, 153 10 Agia Paraskevi Attikis, Athens (Greece)

    2012-04-15

    Highlights: Black-Right-Pointing-Pointer A novel CVD reactor for the developments of double side active TiO{sub 2} membranes. Black-Right-Pointing-Pointer Double side active TiO{sub 2} membranes efficiently photodegrade organic pollutants. Black-Right-Pointing-Pointer A photocatalytic membrane purification device for continuous flow water treatment. - Abstract: A chemical vapour deposition (CVD) based innovative approach was applied with the purpose to develop composite TiO{sub 2} photocatalytic nanofiltration (NF) membranes. The method involved pyrolytic decomposition of titanium tetraisopropoxide (TTIP) vapor and formation of TiO{sub 2} nanoparticles through homogeneous gas phase reactions and aggregation of the produced intermediate species. The grown nanoparticles diffused and deposited on the surface of {gamma}-alumina NF membrane tubes. The CVD reactor allowed for online monitoring of the carrier gas permeability during the treatment, providing a first insight on the pore efficiency and thickness of the formed photocatalytic layers. In addition, the thin TiO{sub 2} deposits were developed on both membrane sides without sacrificing the high yield rates. Important innovation was also introduced in what concerns the photocatalytic performance evaluation. The membrane efficiency to photo degrade typical water pollutants, was evaluated in a continuous flow water purification device, applying UV irradiation on both membrane sides. The developed composite NF membranes were highly efficient in the decomposition of methyl orange exhibiting low adsorption-fouling tendency and high water permeability.

  16. Topical and working papers on heavy water requirements and availability

    International Nuclear Information System (INIS)

    The documents included in this report are: Heavy water requirements and availability; technological infrastructure for heavy water plants; heavy water plant siting; hydrogen and methane availability; economics of heavy water production; monothermal, water fed heavy water process based on the ammonia/hydrogen isotopic exchange; production strategies to meet demand projections; hydrogen availability; deuterium sources; the independent UHDE heavy water process

  17. Chemical elimination of alumina in suspension in nuclear reactors heavy water; Elimination de l'alumine en suspension dans l'eau lourde des reacteurs nucleaires par voie chimique

    Energy Technology Data Exchange (ETDEWEB)

    Ledoux, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1967-02-01

    Corrosion of aluminium in contact with moderating water in nuclear reactor leads to the formation of an alumina hydrosol which can have an adverse effect on the operation of the reactor. Several physical methods have been used in an attempt to counteract this effect. The method proposed here consists in the elimination of the aluminium by dissolution and subsequent fixation in the ionic form on mixed-bed ion-exchange resin. In order to do this, the parameters and the values of these parameters most favorable to the dissolution process have been determined. If the moderator is heavy water, the deuterated acid can be prepared by converting a solution in heavy water to a salt of the acid using a deuterated cationic resin. (author) [French] La corrosion de l'aluminium au contact de l'eau moderatrice des reacteurs nucleaires, donne lieu a la formation d'un hydrosol d'alumine nuisible au bon fonctionnement des reacteurs. Plusieurs methodes physiques ont ete mises en oeuvre pour pallier ces inconvenients. On propose ici d'eliminer l'alumine par solubilisation pour la fixer ensuite sous forme ionique par des resines echangeuses d'ions, en lit melange. A cette fin on determine les parametres et leurs grandeurs favorables a cette solubilisation. Si le moderateur est de l'eau lourde la preparation d'acide deutere peut etre effectuee par passage d'une solution en eau lourde a un sel de l'acide sur resine cationique deuteree.

  18. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  19. Safety of thermal water reactors

    International Nuclear Information System (INIS)

    This book reports on the latest European research into the safety of thermal water reactors, based on the presentation and evaluation of results obtained from research projects undertaken in different national laboratories of the European Community. Information is included under the following areas of research: 1.) The loss of coolant accident (LOCA) and the functioning and performance of the emergency core cooling system; 2.) The protection of nuclear power plants against external gas cloud explosions; and 3.) The release and distribution of radioactive fission products in the atmosphere following a reactor accident

  20. Reactor water level control system

    International Nuclear Information System (INIS)

    A BWR type reactor comprises a control valve disposed in a reactor water draining pipelines and undergoing an instruction to control the opening degree, an operation board having a setting device for generating the instruction and a control board for giving the instruction generated by the setting device to the control valve. The instruction is supplied from the setting device to the control valve by way of a control circuit to adjust the opening degree of the control valve thereby controlling the water level in the reactor. In addition, a controller generating an instruction independent of the setting device and a signal transmission channel for signal-transmitting the instruction independent of the control circuit are disposed, to connect the controller electrically to the signal transmission. The signal transmission channel and the control circuit are electrically connected to the control valve switchably with each other. Since instruction can be given to the control valve even at a periodical inspection or modification when the setting device and the control circuit can not be used, the reactor water level can be controlled automatically. Then, operator's working efficiency upon inspection can be improved remarkably. (N.H.)

  1. Treatemnt of Wastewater with Modified Sequencing Batch Biofilm Reactor Technology

    Institute of Scientific and Technical Information of China (English)

    胡龙兴; 刘宇陆

    2002-01-01

    This paper describes the removel of COD and nitrogen from wastewater with modified sequencing batch biofilm reactor,The strategy of simultaneous feeding and draining was explored.The results show that introduction of a new batch of wastewater and withdrawal of the purifeid water can be conducted simultaneously with the maximum volumetric exchange rate of about 70%,Application of this feeding and draining mode leads to the reduction of the cycle time,the increase of the utilization of the reactor volume and the simplification of the reactor structure.The treatment of a synthetic wastewater containing COD and nitrogen was investigated.The operation mode of F(D)-O(i.e.,simultaneous feeding and draining followed by the aerobic condition)was adopted.It was found that COD was degraded very fast in the initial reaction period of time,then reduced slowly and the ammonia nitrogen and nitrate nitrogen concentrations decreased and increased with time respectively,while the nitrite nitrogen level increased first and then reduced.The relationship between the COD or ammonia nitrogen loading and its removal rate was examined,and the removal of COD,ammonia nitrogen and total nitrogen could exceed 95%,90%and 80% respectively,The fact that nitrogen could e removed more completely under constant aeration(aerobic condition)of the SBBR operation mode is very interesting and could be explained in several respects.

  2. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    International Nuclear Information System (INIS)

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution

  3. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  4. The safety of light water reactors

    International Nuclear Information System (INIS)

    The book describes the principles and practices of reactor safety as applied to the design, regulation and operation of both pressurized water reactors and boiling water reactors. The central part of the book is devoted to methods and results of safety analysis. Some significant events are described, notably the Three Mile Island accident. The book concludes with a chapter on the PIUS principle of inherent reactor safety as applied to the SECURE type of reactor developed in Sweden. (G.B.)

  5. Pressurized water reactor inspection procedures

    International Nuclear Information System (INIS)

    Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.)

  6. Heavy water technology and its contribution to energy sustainability

    International Nuclear Information System (INIS)

    Full text: As the global nuclear industry expands several markets are exploring avenues and technologies to underpin energy security. Heavy water reactors are the most versatile power reactors in the world. They have the potential to extend resource utilization significantly, to allow countries with developing industrial infrastructures access to clean and abundant energy, and to destroy long-lived nuclear waste. These benefits are available by choosing from an array of possible fuel cycles. Several factors, including Canada's early focus on heavy-water technology, limited heavy-industry infrastructure at the time, and a desire for both technological autonomy and energy self-sufficiency, contributed to the creation of the first commercial heavy water reactor in 1962. With the maturation of the industry, the unique design features of the now-familiar product-on-power refuelling, high neutron economy, and simple fuel design-make possible the realization of its potential fuel-cycle versatility. As resource constrains apply pressure on world markets, the feasibility of these options have become more attractive and closer to entering widespread commercial application

  7. Light-Water-Reactor Safety Research Program. Quarterly progress report, January--March 1978

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-05-01

    This progress report summarizes the Argonne National Laboratory work performed during January, February, and March 1978 on water-reactor-safety problems. The following research and development areas are covered: (1) Loss-of-coolant Accident Research: Heat Transfer and Fluid Dynamics; (2) Transient Fuel Response and Fission-product Release Program; (3) Mechanical Properties of Zircaloy Containing Oxygen; and (4) Steam-explosion Studies.

  8. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Production of heavy water. An analysis from export control point of view

    International Nuclear Information System (INIS)

    The main use for heavy water is in heavy water moderated nuclear reactors. This type of reactor is optimal for producing plutonium and therefore countries with nuclear weapons ambitions show interest for such reactors. Most equipment used in heavy water production facilities are under export control to prevent clandestine heavy water production. Heavy water is produced by increasing the concentration of deuterium in the water relative to natural hydrogen. For use in nuclear reactors the deuterium content has to be increased from 0,0155 % to more than 99,75 %. The enrichment can be achieved by several different processes where the GS-process (hydrogen-sulphide exchange) and ammonia-hydrogen exchange process are the two most cost efficient. The energy consumption for these processes is however relatively high. This is due to low separation efficiency which is caused by the relatively small difference in chemical properties between heavy water and natural water. FOI has, under contract work financed by SKI, performed a study of different production processes for heavy water and identified equipment used in the processes. The identified equipment includes both equipment under export control and other sensitive equipment which is important to prevent countries with nuclear weapons ambitions to acquire

  10. Environmentally assisted cracking in light water reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.

    2007-11-06

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2002. Topics that have been investigated include: (a) environmental effects on fatigue crack initiation in carbon and low-alloy steels and austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs in BWRs, (c) evaluation of causes and mechanisms of irradiation-assisted cracking of austenitic SS in PWRs, and (d) cracking in Ni-alloys and welds. A critical review of the ASME Code fatigue design margins and an assessment of the conservation in the current choice of design margins are presented. The existing fatigue {var_epsilon}-N data have been evaluated to define the effects of key material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Experimental data are presented on the effects of surface roughness on fatigue crack initiation in these materials in air and LWR environments. Crack growth tests were performed in BWR environments on SSs irradiated to 0.9 and 2.0 x 10{sup 21} n x cm{sup -2}. The crack growth rates (CGRs) of the irradiated steels are a factor of {approx}5 higher than the disposition curve proposed in NUREG-0313 for thermally sensitized materials. The CGRs decreased by an order of magnitude in low-dissolved oxygen (DO) environments. Slow-strain-rate tensile (SSRT) tests were conducted in high-purity 289 C water on steels irradiated to {approx}3 dpa. The bulk S content correlated well with the susceptibility to intergranular SCC in 289 C water. The IASCC susceptibility of SSs that contain >0.003 wt. % S increased drastically. bend tests in inert environments at 23 C were conducted on broken pieces of SSRT specimens and on unirradiated specimens of the same materials after hydrogen charging. The results of the tests and a review of other data in the literature

  11. Feasibility study on the thorium fueled boiling water breeder reactor

    International Nuclear Information System (INIS)

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  12. The Advanced Light Water Reactor

    International Nuclear Information System (INIS)

    The U. S. Advanced Light Water Reactor Program is a forward-looking program designed to produce viable nuclear generating system candidates to meet the very real, and perhaps imminent, need for new power generation capacity in the U. S. and around the world. The ALRR Program is an opportunity to move ahead with confidence, to confront problems today which must be confronted if the U. S. electrical utilities are to continue to meet their commitment to provide safe, reliable, economical electrical power to the nation in the years ahead. Light water reactor technology is today playing a vital role in the production of electricity to meet the world's needs. At present about 13% of the world's electricity is supplied by nuclear power plants, most of those light water reactors. Nevertheless, there is a clear need for expanded use of nuclear generation. Here in Korea and elsewhere in Asia, demand for electricity has continued to increase at a very high rate. In the United States demand growth has been more moderate, but a large number of existing stations will be ready for replacement in the next two decades, and all countries face the problem of dwindling fuel supplies and growing environmental impact of fossil-fired power plants. Despite the evident need for expanded nuclear generation capacity in the United States, there have been no new plants ordered in the past ten years and at present there are no immediate prospects for new plant orders. Concerns about safety, the high cost of recent nuclear stations, and the current excess of electrical generation capacity in the United States, have combined to interrupt completely the growth of this vital power supply system

  13. Water chemistry management of research reactor in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yoshijima, Tetsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    The JRR-3M cooling system consists of four systems, namely; (1) primary cooling system, (2) heavy water cooling system, (3) helium system and (4) secondary cooling system. The heavy water is used for reflector and pressurized with helium gas. Water chemistry management of the JRR-3M cooling systems is one of the important subject for the safety operation. The main objects are to prevent the corrosion of cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the generation of radioactive waste. All measured values were within the limits of specifications and JRR-3M reactor was operated with safety in 1996. Spent fuels of JRR-3M reactor are stored in the spent fuel pool. This pool water has been analyzed to prevent corrosion of aluminum cladding of spent fuels. Water chemistry of spent fuel pool water is applied to the prevention of corrosion of aluminum alloys including fuel cladding. The JRR-2 reactor was eternally stopped in December 1996 and is now under decommissioning. The JRR-2 reactor is composed of heavy water tank, fuel guide tube and horizontal experimental hole. These are constructed of aluminum alloy and biological shield and upper shield are constructed of concrete. Three types of corrosion of aluminum alloy were observed in the JRR-2. The Alkaline corrosion of aluminum tube occurred in 1972 because of the mechanical damage of the aluminum fuel guide tube which is used for fuel handling. Modification of the reactor top shield was started in 1974 and completed in 1975. (author)

  14. Recycle of LWR [Light Water Reactor] actinides to an IFR [Integral Fast Reactor

    International Nuclear Information System (INIS)

    A large quantity of actinide elements is present in irradiated Light Water Reactor (LWR) fuel that is stored throughout the world. Because of the high fission-to-capture ratio for the transuranium (TRU) elements with the high-energy neutrons in the metal-fueled Integral Fast Reactor (IFR), that reactor can consume these elements effectively. The stored fuel represents a valuable resource for an expanding application of fast power reactors. In addition, removal of the TRU elements from the spent LWR fuel has the potential for increasing the capacity of a high-level waste facility by reducing the heat loads and increasing the margin of safety in meeting licensing requirements. Argonne National Laboratory (ANL) is developing a pyrochemical process, which is compatible with the IFR fuel cycle, for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. The major objective of the LWR fuel recovery process is high TRU element recovery, with decontamination a secondary issue, because fission product removal is accomplished in the IFR process. The extensive pyrochemical processing studies of the 1960s and 1970s provide a basis for the design of possible processes. Two processes were selected for laboratory-scale investigation. One is based on the Salt Transport Process studied at ANL for mixed-oxide fast reactor fuel, and the other is based on the blanket processing studies done for ANL's second Experimental Breeder Reactor (EBR-2). This paper discusses the two processes and is a status report on the experimental studies. 5 refs., 2 figs., 2 tabs

  15. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  16. European simplified boiling water reactor (ESBWR) plant

    International Nuclear Information System (INIS)

    This paper covers innovative ideas which made possible the redesign of the US 660-MW Simplified Boiling Water Reactor (SBWR) Reactor Island for a 1,200-MW size reactor while actually reducing the building cost. This was achieved by breaking down the Reactor Island into multiple buildings separating seismic-1 from non-seismic-1 areas, providing for better space utilization, shorter construction schedule, easier maintainability and better postaccident accessibility

  17. Modified divergence theorem for analysis and optimization of wall reflecting cylindrical UV reactor

    Directory of Open Access Journals (Sweden)

    Milanović Đurđe R.

    2011-01-01

    Full Text Available In this paper, Modified Divergence Theorem (MDT, known in earlier literature as Gauss-Ostrogradsky theorem, was formulated and proposed as a general approach to electromagnetic (EM radiation, especially ultraviolet (UV radiation reactor modeling. Formulated mathematical model, based on MDT, for multilamp UV reactor was applied to all sources in a reactor in order to obtain intensity profiles at chosen surfaces inside reactor. Applied modification of MDT means that intensity at a real opaque or transparent surface or through a virtual surface, opened or closed, from different sides of the surface are added and not subtracted as in some other areas of physics. Derived model is applied to an example of the multiple UV sources reactor, where sources are arranged inside a cylindrical reactor at the coaxial virtual cylinder, having the radius smaller than the radius of the reactor. In this work, optimization of a reactor means maximum transfer of EM energy sources into the fluid for given fluid absorbance and fluid flow-dose product. Obtained results, for in advanced known water quality, gives unique solution for an optimized model of a multilamp reactor geometry. As everyone can easily verify, MDT is very good starting point for every reactor modeling and analysis.

  18. Thermophysical properties of materials for water cooled reactors

    International Nuclear Information System (INIS)

    The IAEA Co-ordinated Research Programme (CRP) to establish a thermophysical properties data base for light and heavy water reactor materials was organized within the framework of the IAEA's International Working Group on Advanced Technologies for Water Cooled Reactors. The work within the CRP started in 1990. The objective of the CRP was to collect and systemaize a thermophysical properties data base for light and heavy water reactor materials under normal operating, transient and accident conditions. The important thermophysical properties include thermal conductivity, thermal diffusivity, specific heat capacity, enthalpy, thermal expansion and others. These properties as well as the oxidation of zirconium-based alloys, the thermophysical characteristics of high temperature concrete-core melt interaction and the mechanical properties of construction materials are presented in this report. It is hoped that this report will serve as a useful source of thermophysical properties data for water cooled reactor analyses. The properties data are maintained on the THERSYST system at the University of Stuttgart, Germany and are internationally available. Refs, figs, tabs

  19. Water issues associated with heavy oil production.

    Energy Technology Data Exchange (ETDEWEB)

    Veil, J. A.; Quinn, J. J.; Environmental Science Division

    2008-11-28

    Crude oil occurs in many different forms throughout the world. An important characteristic of crude oil that affects the ease with which it can be produced is its density and viscosity. Lighter crude oil typically can be produced more easily and at lower cost than heavier crude oil. Historically, much of the nation's oil supply came from domestic or international light or medium crude oil sources. California's extensive heavy oil production for more than a century is a notable exception. Oil and gas companies are actively looking toward heavier crude oil sources to help meet demands and to take advantage of large heavy oil reserves located in North and South America. Heavy oil includes very viscous oil resources like those found in some fields in California and Venezuela, oil shale, and tar sands (called oil sands in Canada). These are described in more detail in the next chapter. Water is integrally associated with conventional oil production. Produced water is the largest byproduct associated with oil production. The cost of managing large volumes of produced water is an important component of the overall cost of producing oil. Most mature oil fields rely on injected water to maintain formation pressure during production. The processes involved with heavy oil production often require external water supplies for steam generation, washing, and other steps. While some heavy oil processes generate produced water, others generate different types of industrial wastewater. Management and disposition of the wastewater presents challenges and costs for the operators. This report describes water requirements relating to heavy oil production and potential sources for that water. The report also describes how water is used and the resulting water quality impacts associated with heavy oil production.

  20. Technical status study of heavy water enrichment

    International Nuclear Information System (INIS)

    Technical status study of heavy water enrichment in Indonesia and also in the world has been done. Heavy water enrichment processes have been investigated were water distillation, hydrogen distillation, laser enrichment, electrolysis and isotop exchange. For the isotop exchange, the chemical pair can be used were water-hydrogen sulphite, ammonium-hydrogen, aminomethane-hydrogen, and water-hydrogen. For the isotope exchange, there was carried out by mono thermal or bi thermal. The highest producer of heavy water is Canada, and the other producer is USA, Norwegian and India. The processes be used in the world are isotope exchange Girdler Sulphide (GS), distillation and electrolysis. Research of heavy water carried out in Batan Yogyakarta, has a purpose to know the characteristic of heavy water purification. Several apparatus which has erected were 3 distillation column: Pyrex glass of 2 m tall, stainless steel column of 3 m tall and steel of 6 m tall. Electrolysis apparatus is 50 cell electrolysis and an isotope exchange unit which has catalyst: Ni- Cr2O3 and Pt-Carbon. These apparatus were not ready to operate. (author)

  1. Study on Recycle of Materials and Components From Waste Streams During Decommissioning for Heavy Water Research Reactor%重水研究堆退役废物再利用研究

    Institute of Scientific and Technical Information of China (English)

    岳维宏; 逄锦鑫

    2013-01-01

    实现废物再利用是废物最小化的重要措施之一,从废物流中将有潜在利用价值的物料分离出来实现再利用可大幅减少对环境的影响。本文以中国原子能科学研究院重水研究堆退役为实例研究了放射性废物再利用问题。通过全面分析和计算重水研究堆在退役期间产生的各类废物,得出具有一定数量的物料有潜在的利用价值,可直接或经适当处理后再利用在其他行业领域中。研究表明,通过采取废物最小化控制措施(如废物分类和废物流分离等),采用适当的去污技术和执行清洁解控要求,至少可使重水研究堆退役过程中产生的几十吨钢铁、10 t铝材和5 t重水实现再利用。%The recycle of valuable materials from potential waste streams is one of important elements of waste minimization ,and it can minimize the environment impact . The recycle of the arising was researched with taking the decommissioning of heavy water research reactor (HWRR) in China Institute of Atomic Energy as an example .By analyzing all the possible wastes that could generate during the decommissioning of HWRR ,some amount of materials have potential values to recycle and may be used either directly or after appropriate treatment for other purposes .The research results show that in HWRR decommissioning at least tons of irons ,10 tons of aluminum and 5 tons of heavy water can be recycled by carrying out the waste minimization control measures (eg .waste classification and waste stream segregation) ,adopting appropriate decontamination technologies ,and performing the requirements of clearance .

  2. TA-2 Water Boiler Reactor Decommissioning Project

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, M.E. (ed.); Montoya, G.M.

    1991-06-01

    This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m{sup 3} of low-level solid radioactive waste and 35 m{sup 3} of mixed waste. 15 refs., 25 figs., 3 tabs.

  3. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  4. Reactor Safety Commission Code of Practice for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

  5. Topical and working papers on heavy water accountability and safeguards

    International Nuclear Information System (INIS)

    This report contains the following papers: 1) Statement of IAEA concerning safeguarding of heavy water; 2) Preliminary Canadian Comments on IAEA document on heavy water safeguards; 3) Heavy water accountability 03.10.78; 4) Heavy water accountability 05.04.79

  6. Maintenance technologies for degradation of pressurized water reactor power plants

    International Nuclear Information System (INIS)

    As a countermeasure against SCC (stress corrosion cracking), MHI (Mitsubishi Heavy Industries, Ltd.) have developed some residual stress improvement methods, as Water Jet Peening (WJP) for components under water condition, and Shot Peening by Ultrasonic-wave vibration (USP) for components under air condition. The SCC occurred in high nickel based metal and welding material in pressurized water reactor (PWR) plants has become to be conspicuous issue in both Japan and abroad. In this paper, validity of stress improvement by WJP/USP for SCC mitigation has been verified for area with small cracks. (author)

  7. Pulse radiolysis studies of liquid heavy water at temperatures up to 250 degrees C

    Energy Technology Data Exchange (ETDEWEB)

    Stuart, C.R.; Ouellette, D.C.; Elliot, A.J

    2002-09-01

    This report documents the rate constants and associated activation energies for the reactions of the primary radical species, e{sub aq}{sup -}, {center_dot}OD and {center_dot}D, which are formed during the radiolysis of heavy water within the temperature range 20 to 250 {sup o}C. These heavy-water data have been compared with the corresponding information for light water. These kinetic data form part of the database that is required to model the aqueous radiation chemistry that occurs within the core of the heavy water cooled and moderated CANDU reactor. (author)

  8. Light water reactor safety research project

    International Nuclear Information System (INIS)

    The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

  9. Modified Mathematical Model For Neutralization System In Stirred Tank Reactor

    Directory of Open Access Journals (Sweden)

    Ahmmed Saadi Ibrehem

    2011-05-01

    Full Text Available A modified model for the neutralization process of Stirred Tank Reactors (CSTR reactor is presented in this study. The model accounts for the effect of strong acid [HCL] flowrate and strong base [NaOH] flowrate with the ionic concentrations of [Cl-] and [Na+] on the Ph of the system. In this work, the effect of important reactor parameters such as ionic concentrations and acid and base flowrates on the dynamic behavior of the CSTR is investigated and the behavior of mathematical model is compared with the reported models for the McAvoy model and Jutila model. Moreover, the results of the model are compared with the experimental data in terms of pH dynamic study. A good agreement is observed between our model prediction and the actual plant data. © 2011 BCREC UNDIP. All rights reserved(Received: 1st March 2011, Revised: 28th March 2011; Accepted: 7th April 2011[How to Cite: A.S. Ibrehem. (2011. Modified Mathematical Model For Neutralization System In Stirred Tank Reactor. Bulletin of Chemical Reaction Engineering & Catalysis, 6(1: 47-52. doi:10.9767/bcrec.6.1.825.47-52][How to Link / DOI: http://dx.doi.org/10.9767/bcrec.6.1.825.47-52 || or local:  http://ejournal.undip.ac.id/index.php/bcrec/article/view/825 ] | View in 

  10. Functional systems of a pressurized water reactor

    International Nuclear Information System (INIS)

    The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

  11. Efficient Water Management in Water Cooled Reactors

    International Nuclear Information System (INIS)

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world'. One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property.' The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. Water scarcity is becoming one of the most pressing crises affecting the planet. A reliable supply of water and energy is an important prerequisite for sustainable development. A large number of nuclear power reactors are being planned in many developing countries to address these countries' increasing energy demands and their limited fossil resources. New construction is expected in the USA, Europe and Asia, as well. Reducing water use and consumption by nuclear power plants is likely to help developing countries in introducing nuclear power into their energy supply mix. A large

  12. Ageing management in heavy water plants

    International Nuclear Information System (INIS)

    With proper design and construction ageing can be minimized. It is important to understand the mechanics of ageing specific to service, develop baseline data and monitor to ensure that there are no premature failures especially where the service conditions are extreme and media used is highly corrosive and hazardous such as in heavy water plants. The key lies in an effective in-service inspection and determination of residual life for decision making vis-a-vis upgrading and ensuring safety. While quite a bit of work in this direction has been done in the Heavy Water Board, a lot more ground needs to be covered. (author)

  13. 基体改进石墨炉原子吸收测定净水器自制水中的重金属%Determination of Heavy Metals in Homemade Water of Water Purifier by Graphite Furnace Atomic Absorption Spectrometry with Matrix Modifier

    Institute of Scientific and Technical Information of China (English)

    康凌之; 俞卫平; 王聪

    2016-01-01

    建立了石墨炉原子吸收分光光度法测定净水器自制水中重金属含量的方法,为水中重金属的测定提供可靠的途径。较系统地研究了基体改进剂磷酸二氢铵、硝酸镍、硝酸镁在测定净水器自制水中铬、镉、硒、镍元素的运用,并讨论了基体改进剂消除干扰的作用机理,同时确定了最佳的仪器测试条件。本方法用于测定净水器中的痕量元素,结果令人满意。%This study had established the method of heavy metal content in water of water purifier by graphite furnace atomic absorption spectrophotometry. It provided a reliable way for the determination of heavy metals in water. The use of ammonium dihydrogen phosphate, nickel nitrate, magnesium nitrate as matrix modifier in determination of chromium, cadmium, selenium, nickel in homemade water of water purifi-er has been systematically studied. The mechanism of their eliminating effects was discussed. Meanwhile the best instrument test conditions were determined. Appling this method to analyze the elements in homemade water, the results were satisfactory.

  14. Development of integrated modular water reactor

    International Nuclear Information System (INIS)

    In order to adapt for environmental problem to reduce emission of greenhouse effect gas and develop a power generation plant with economical efficiency and small output, a number of R and Ds on small scale reactors have been progressed without any practice. The largest subject on the development consists in how cost of construction, operation and maintenance on the small scale reactors can be reduced to those of large scale ones by its specific technology. Therefore, if by adding wide simplification of apparatuses based on introduction of novel technology and added values specific to the small scale reactors, a business model on installing many small scale reactors can be established, a practicable feasibility of the small scale reactors valuable to introduction of actual scale machine will enable to be found. Authors have progressed development on a novel small scale reactor capable of flexibly corresponding to social needs for nuclear power generation. As a result, by integrating the reactor system and introducing self pressurisation and natural circulation as an innovative 300,000 kW output class small scale reactor, a plant concept reducing feasibility to occur any large scale accident to its ultimate limit together with planning wide simplification of apparatus could be established. The reactor is the titled integrated modular water reactor (IMR), and is at present under investigation on its formability confirmation and its concept design. Here were reported on plant concepts and characteristics of IMR, in this report. (G.K.)

  15. Modified accumulation of selected heavy metals in Bt transgenic rice

    Institute of Scientific and Technical Information of China (English)

    WANG Haiyan; HUANG Jianzhong; YE Qingfu; WU Dianxing; CHEN Ziyuan

    2009-01-01

    Safety assessment of genetically modified crops generally does not take into account the potential hazard of altered patterns of heavy metal accumulation in plants.A pot experiment was conducted under greenhouse conditions to evaluate the impact of heavy metal amendments on the accumulation of Cd,Cu,Pb and Zn in a Bt transgenic rice Ke-Ming-Dao (KMD) and its wild-type Xiushui 11 (Xs11).In control soils,significant difference was only found in contents of Cu (p < 0.01) and Pb (p < 0.05) in straw between KMD and Xs11.At three levels of Cd amendments (5,10,and 20 mg/kg),the Cd contents in grain and straw of KMD were significantly higher than those of Xs11,and all grain Cd contents were significantly higher than the international criteria (0.2-0.4 mg/kg) as specified by the Codex Alimentarius Commission (CAC).These results implied that it may be unsafe for growing Bt transgenic rice in heavily Cd-polluted areas.No significant difference in Zn was found between the two varieties with the exception of roots at Zn amendment level of 600 mg/kg,while Pb contents in KMD were much higher in the straw at the lead amendment level of 1000 mg/kg and in the root at 250 mg Pb/kg.Data on the heavy metal accumulation patterns for the genetically modified rice may be used for the selection of growing areas as well as for plant residue management for Bt rice.

  16. Self-Sustaining Thorium Boiling Water Reactors

    OpenAIRE

    Ehud Greenspan; Jasmina Vujic; Francesco Ganda; Arias, Francisco J.

    2012-01-01

    A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR) proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorber...

  17. Chemically-Modified Cellulose Paper as a Microstructured Catalytic Reactor

    Directory of Open Access Journals (Sweden)

    Hirotaka Koga

    2015-01-01

    Full Text Available We discuss the successful use of chemically-modified cellulose paper as a microstructured catalytic reactor for the production of useful chemicals. The chemical modification of cellulose paper was achieved using a silane-coupling technique. Amine-modified paper was directly used as a base catalyst for the Knoevenagel condensation reaction. Methacrylate-modified paper was used for the immobilization of lipase and then in nonaqueous transesterification processes. These catalytic paper materials offer high reaction efficiencies and have excellent practical properties. We suggest that the paper-specific interconnected microstructure with pulp fiber networks provides fast mixing of the reactants and efficient transport of the reactants to the catalytically-active sites. This concept is expected to be a promising route to green and sustainable chemistry.

  18. Bubble column reactor fluid dynamic study at pilot plant scale for residue and extra heavy crude oil upgrading technology

    Energy Technology Data Exchange (ETDEWEB)

    Sardella, R.; Medina, H. [Infrastructure and Upgrading Department PDVSA-Intevep (Venezuela); Zacarias, L.; Paiva, M. [Refining Department. PDVSA-Intevep (Venezuela)

    2011-07-01

    Bubble column reactors are used in several applications because of their simplicity and low cost; a new technology was developed to convert extra heavy crude oil into upgraded crude using a bubble column reactor. To design this kind of reactor, a lot of parameters like flow regime, gas hold up and dispersion coefficient have to be taken into account. This study aimed at determining the fluid dynamic behaviour of a bubble column working under Aquaconversion operating conditions. Experiments were undertaken on air-tap water and air-light oil systems under atmospheric conditions with various gas superficial velocities and liquid flowrates. Results showed that gas hold up increases with superficial gas velocity but is independent of liquid flowrate and that both systems tested work at the same flow regimes. This paper showed that under the experimental conditions used, this reactor tends to be a complete mixing reactor.

  19. Report on the workshop on atomic and plasma physics requirements for heavy ion fusion, Argonne National Lab., December 13-14, 1979

    International Nuclear Information System (INIS)

    Atomic, molecular, and plasma physics areas that are relevant to inertial confinement fusion by energetic heavy ions are identified. Discussions are confined to problems related to the design of heavy ion accelerators, accumulation of ions in storage rings, and the beam transport in a reactor vessel

  20. Measurement of the purity of graphite and heavy water

    International Nuclear Information System (INIS)

    The analytical methods used by the C.E.A. are described, I -- Graphite. The determination of the change in the neutron capture cross section from sample to sample is determined by, an oscillation method in the Zoe reactor, or by measuring the attenuation of a neutron flux in the subcritical system Mireille. Methods of analysing total ash, B, H, Cl, Na, Ca. Fe, Mo, Ti, V, Sm, Eu, Dy, S, Co and Cd are described and mean results are given. The methods for sampling are indicated. II -- Heavy crater. The isotopic analysis of heavy water is carried out by infra-red absorption measurements. Chemical purity is evaluated by electrical conductivity measurements, B, Na, Mg, K, Cr, Mn, Ni, Cu, Cd, are determined by spectrographic methods, and Cl-, NO3-, SO4--, NH4+ by chemical methods; finally, sensitive pH measurements are described

  1. Inherently safe light water reactors

    International Nuclear Information System (INIS)

    Today's large nuclear power reactors of world-wise use have been designed based on the philosophy. It seems that recent less electricity demand rates, higher capital cost and the TMI accident let us acknowledge relative small and simplified nuclear plants with safer features, and that Chernobyl accident in 1983 underlines the needs of intrinsic and passive safety characteristics. In such background, several inherently safe reactor concepts have been presented abroad and domestically. First describing 'Can inherently safe reactors be designed,' then I introduce representative reactor concepts of inherently safe LWRs advocated abroad so far. All of these innovative reactors employ intrinsic and passive features in their design, as follows: (1) PIUS, an acronym for Process Inherent Ultimate Safety, or an integral PWR with passive heat sink and passive shutdown mechanism, advocated by ASEA-ATOM of Sweden. (2) MAP(Minimum Attention Plant), or a self-pressurized, natural circulation integral PWR, promoted by CE Inc. of the U.S. (3) TPS(TRIGA Power System), or a compact PWR with passive heat sink and inherent fuel characteristics of large prompt temperature coefficient, prompted by GA Technologies Inc. of the U.S. (4) PIUS-BWR, or an inherently safe BWR employing passively actuated fluid valves, in competition with PIUS, prompted by ORNL of the U.S. Then, I will describe the domestic trends in Japan and the innovative inherently safe LWRs presented domestically so far. (author)

  2. Canadian heavy water production - 1970 to 1980

    International Nuclear Information System (INIS)

    In the last decade, heavy water production in Canada has progressed from the commissioning of a single unit plant in Nova Scotia to a major production industry employing 2200 persons and operating three plants with an aggregate annual production capability in excess of 1800 Mg. The decade opened with an impending crisis in the supply of heavy water due to failure of the first Glace Bay Heavy Water Plant and difficulty in commissioning the second Canadian plant at Port Hawkesbury. Lessons learned at this latter plant were applied to the Bruce plant where the first two units were under construction. When the Bruce units were commissioned in 1973 the rate of approach to design production rates was much improved, renewing confidence in Canada's ability to succeed in large scale heavy water production. In the early 1970's a decision was made to rehabilitate the Glace Bay plant using a novel flowsheet and this rebuilt plant commenced production in 1976. The middle of the decade was marked by two main events: changes in ownership of the operating plants and initiation of a massive construction program to support the forecast of a rapidly expanding CANDU power station construction program. New production units embodying the best features of their predecessors were committed at Bruce by Ontario Hydro and at La Prade, Quebec, by AECL. The high growth rate in electrical demand did not continue and some new plant construction was curtailed. The present installed production capacity will now probably be adequate to meet anticipated demand for the next decade. Canadian plants have now produced more than 7800 Mg of heavy water at a commercially acceptable cost and with a high degree of safety and compliance with appropriate environmental regulations

  3. Heavy ion beam transport through liquid lithium first wall ICF reactor cavities

    International Nuclear Information System (INIS)

    This analysis addresses the critical issue of the final transport of a heavy ion beam in an inertial confinement fusion reactor. The beam must traverse the reaction chamber from the final focusing lens to the target without being disrupted. This requirement has a strong impact on the reactor design. It is essential to the development of ICF fusion reactor technology, that the restrictions placed on the reactor engineering parameters by final beam transport consideration be understood early on

  4. Pressurized water reactor flow skirt apparatus

    Science.gov (United States)

    Kielb, John F.; Schwirian, Richard E.; Lee, Naugab E.; Forsyth, David R.

    2016-04-05

    A pressurized water reactor vessel having a flow skirt formed from a perforated cylinder structure supported in the lower reactor vessel head at the outlet of the downcomer annulus, that channels the coolant flow through flow holes in the wall of the cylinder structure. The flow skirt is supported at a plurality of circumferentially spaced locations on the lower reactor vessel head that are not equally spaced or vertically aligned with the core barrel attachment points, and the flow skirt employs a unique arrangement of hole patterns that assure a substantially balanced pressure and flow of the coolant over the entire underside of the lower core support plate.

  5. Thermohydraulic analysis of pressurized water reactors

    International Nuclear Information System (INIS)

    The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor)

  6. Sorption of Heavy Metal and Organic Pollutants on Modified Soils

    Institute of Scientific and Technical Information of China (English)

    MENG Zhao-Fu; ZHANG Yi-Ping; WANG Guo-Dong

    2007-01-01

    Sorption characteristics of both an organic pollutant(phenol)and a heavy metal(cadmium ion)on the clay layer of a Lou soil(Eum-orthic Anthrosol in Chinese Soil Taxonomy)along with the sorption mechanism were investigated using three soil treatments:modification with a cationic surfactant cetyltrimethylammonium bromide added at an amount equivalent to 50%and 100%of the soil CEC (50%CB and 100%CB),modification with an amphoteric surface-modifying agent dodecyldimethylbetaine(commercially known as BS-12)added at an amount equivalent to 50%and 100%of the soil CEC(50%BS and 100%BS),and an unmodified control(CK).Results showed that the BS soil treatments increased sorption of both the heavy metal Cd2+ and the organic pollutant phenol.The equilibrium sorption amount of Cd2+decreased in the order:50%BS>100%BS>CK>50%CB>100%CB,with the BS soil treatments being about 1.3 to 1.8 times higher and the CB soil treatments about 23%to 41%lower than CK.Both the single-site and two-site Langmuir models could be applied to describe the sorption of Cd2+ in each soil treatment.The equilibrium sorption amount of phenol on the soil samples decreased in the order:100%CB>50%CB>100%BS>50%BS>CK,with the CB soil treatments being 41.0 to 79.6 times higher and the BS soil treatments 4.0 to 8.3 times higher than CK.The Freundlich equation could also be used to describe the sorption characteristics of phenol.In the BS soil treatments,both an organophobic long carbon chain and hydrophilic charged groups resulted in a relatively strong sorption ability for both heavy metals and organic pollutants.In addition.the sorption ratio K,the ratio of phenol sorption amount of the modified soil to that of CK,increased initially and decreased later with the amount of phenol added,and the critical sorption ratio Kc,the peak value of the sorption ratio curve plotted against the added phenol concentration.was a good index for evaluating the sorption ability of phenol in the soil.

  7. Coolant technology of water cooled reactors. V. 1: Chemistry of primary coolant in water cooled reactors

    International Nuclear Information System (INIS)

    This report is a summary of the work performed within the framework of the Coordinated Research Programme on Investigations on Water Chemistry Control and Coolant Interaction with Fuel and Primary Circuit Materials in Water Cooled Power Reactors organized by the IAEA and carried out from 1987 to 1991. It is the continuation of a programme entitled Reactor Water Chemistry Relevant to Coolant-Cladding Interaction (IAEA-TECDOC-429), which ran from 1981 to 1986. Subsequent meetings resulted in the title of the programme being changed to Coolant Technology of Water Cooled Reactors. The results of this Coordinated Research Programme are published in four volumes with an overview in the Technical Reports Series. The titles of the volumes are: Volume 1: Chemistry of Primary Coolant in Water Cooled Reactors; Volume 2: Corrosion in the Primary Coolant Systems of Water Cooled Reactors; Volume 3: Activity Transport Mechanisms in Water Cooled Reactors; Volume 4: Decontamination of Water Cooled Reactors. These publications should be of interest to experts in water chemistry at nuclear power plants, experts in engineering, fuel designers, research and development institutes active in the field and to consultants to these organizations. Refs, figs and tabs

  8. Flow-induced vibration for light water reactors. Final progress report, July 1981-September 1981

    International Nuclear Information System (INIS)

    Flow-Induced Vibration for Light Water Reactors (FIV for LWRs) is a program designed to improve the FIV performance of light water reactors through the development of design criteria, analytical models for predicting behavior of components, and general scaling laws to improve the accuracy of reduced-scale tests, and through 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. This progress report summarizes the accomplishments achieved during the final period from July 1981 to September 1981. This is the last quarterly progress report to be issued for this program

  9. Thermophysical properties of saturated light and heavy water for Advanced Neutron Source applications

    Energy Technology Data Exchange (ETDEWEB)

    Crabtree, A.; Siman-Tov, M.

    1993-05-01

    The Advanced Neutron Source is an experimental facility being developed by Oak Ridge National Laboratory. As a new nuclear fission research reactor of unprecedented flux, the Advanced Neutron Source Reactor will provide the most intense steady-state beams of neutrons in the world. The high heat fluxes generated in the reactor [303 MW(t) with an average power density of 4.5 MW/L] will be accommodated by a flow of heavy water through the core at high velocities. In support of this experimental and analytical effort, a reliable, highly accurate, and uniform source of thermodynamic and transport property correlations for saturated light and heavy water were developed. In order to attain high accuracy in the correlations, the range of these correlations was limited to the proposed Advanced Neutron Source Reactor`s nominal operating conditions. The temperature and corresponding saturation pressure ranges used for light water were 20--300{degrees}C and 0.0025--8.5 MPa, respectively, while those for heavy water were 50--250{degrees}C and 0.012--3.9 MPa. Deviations between the correlation predictions and data from the various sources did not exceed 1.0%. Light water vapor density was the only exception, with an error of 1.76%. The physical property package consists of analytical correlations, SAS codes, and FORTRAN subroutines incorporating these correlations, as well as an interactive, easy-to-use program entitled QuikProp.

  10. Materials for advanced water cooled reactors

    International Nuclear Information System (INIS)

    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

  11. Environmentally assisted cracking in light water reactors. Semiannual progress report, January 1996--June 1996

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E. [and others

    1997-05-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1996 to June 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288{degrees}C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in air and high-purity, low-DO water. 83 refs., 60 figs., 14 tabs.

  12. Environmentally assisted cracking in Light Water Reactors: Semiannual report, April 1993--September 1993. Volume 17

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Karlsen, T.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

    1994-06-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) during the six months from April 1993 to September 1993. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels; (b) EAC of cast stainless steels (SSs); and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions in simulated boiling-water reactor (BWR) water at 289{degree}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section 11 of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

  13. Environmentally assisted cracking in light water reactors. Semiannual report, April 1994--September 1994, Volume 19

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J. [and others

    1995-09-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors from April to September 1994. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in piping and reactor pressure vessels, (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests have been conducted on A106-Gr B and A533-Gr B steels in oxygenated water to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack growth data were obtained on fracture-mechanics specimens of SSs and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289{degrees}C. The data were compared with predictions from crack growth correlations developed at ANL for SSs in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  14. Environmentally assisted cracking in light water reactors. Semiannual report July 1996--December 1996

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J. [Argonne National Lab., IL (United States)] [and others

    1997-10-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, (c) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated BWR water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from a low-carbon content heat of Alloy 600 in high-purity oxygenated water at 289 C. Residual stresses and stress intensity factors were calculated for BWR core shroud welds.

  15. Environmentally assisted cracking in Light Water Reactors: Semiannual report, April 1993--September 1993. Volume 17

    International Nuclear Information System (INIS)

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) during the six months from April 1993 to September 1993. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels; (b) EAC of cast stainless steels (SSs); and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions in simulated boiling-water reactor (BWR) water at 289 degree C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section 11 of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy

  16. Numerical study of the effects of lamp configuration and reactor wall roughness in an open channel water disinfection UV reactor.

    Science.gov (United States)

    Sultan, Tipu

    2016-07-01

    This article describes the assessment of a numerical procedure used to determine the UV lamp configuration and surface roughness effects on an open channel water disinfection UV reactor. The performance of the open channel water disinfection UV reactor was numerically analyzed on the basis of the performance indictor reduction equivalent dose (RED). The RED values were calculated as a function of the Reynolds number to monitor the performance. The flow through the open channel UV reactor was modelled using a k-ε model with scalable wall function, a discrete ordinate (DO) model for fluence rate calculation, a volume of fluid (VOF) model to locate the unknown free surface, a discrete phase model (DPM) to track the pathogen transport, and a modified law of the wall to incorporate the reactor wall roughness effects. The performance analysis was carried out using commercial CFD software (ANSYS Fluent 15.0). Four case studies were analyzed based on open channel UV reactor type (horizontal and vertical) and lamp configuration (parallel and staggered). The results show that lamp configuration can play an important role in the performance of an open channel water disinfection UV reactor. The effects of the reactor wall roughness were Reynolds number dependent. The proposed methodology is useful for performance optimization of an open channel water disinfection UV reactor.

  17. Numerical study of the effects of lamp configuration and reactor wall roughness in an open channel water disinfection UV reactor.

    Science.gov (United States)

    Sultan, Tipu

    2016-07-01

    This article describes the assessment of a numerical procedure used to determine the UV lamp configuration and surface roughness effects on an open channel water disinfection UV reactor. The performance of the open channel water disinfection UV reactor was numerically analyzed on the basis of the performance indictor reduction equivalent dose (RED). The RED values were calculated as a function of the Reynolds number to monitor the performance. The flow through the open channel UV reactor was modelled using a k-ε model with scalable wall function, a discrete ordinate (DO) model for fluence rate calculation, a volume of fluid (VOF) model to locate the unknown free surface, a discrete phase model (DPM) to track the pathogen transport, and a modified law of the wall to incorporate the reactor wall roughness effects. The performance analysis was carried out using commercial CFD software (ANSYS Fluent 15.0). Four case studies were analyzed based on open channel UV reactor type (horizontal and vertical) and lamp configuration (parallel and staggered). The results show that lamp configuration can play an important role in the performance of an open channel water disinfection UV reactor. The effects of the reactor wall roughness were Reynolds number dependent. The proposed methodology is useful for performance optimization of an open channel water disinfection UV reactor. PMID:27108375

  18. Hydrothermally modified fly ash for heavy metals and dyes removal in advanced wastewater treatment

    International Nuclear Information System (INIS)

    Fly ash resulted from coal burning is a waste that can be used in wastewater treatment for removal of dyes and heavy metals by adsorption. Class “F” fly ash (FA), collected from the Central Heat and Power (CHP) Plant Brasov (Romania), with oxides composition SiO2/Al2O3 over 2.4 was used for obtaining a new substrate with good adsorption capacity for dyes and heavy metals from wastewater. A new material was obtained from modified fly ash with NaOH and hexadecyltrimethylammonium bromide (HTAB) a cationic surfactant. Contact time, optimum amount of substrate and the pH corresponding to 50 mL solution of pollutants were the parameters optimized for obtaining the maximum efficiency in the adsorption process. The optimized adsorption parameters were further used in thermodynamic and kinetic studies of the adsorption processes. The adsorption kinetic mechanisms, and the substrate capacities are further discussed correlated with the surface structure (XRD), composition (EDS, FTIR), and morphology (SEM, AFM). The results indicate that the novel nano-substrate composite with fly ash modified can be used as an efficient and low cost adsorbent for simultaneous removal of dyes and heavy metals, the resulted water respects the discharge regulations.

  19. Hydrothermally modified fly ash for heavy metals and dyes removal in advanced wastewater treatment

    Science.gov (United States)

    Visa, Maria; Chelaru, Andreea-Maria

    2014-06-01

    Fly ash resulted from coal burning is a waste that can be used in wastewater treatment for removal of dyes and heavy metals by adsorption. Class “F” fly ash (FA), collected from the Central Heat and Power (CHP) Plant Brasov (Romania), with oxides composition SiO2/Al2O3 over 2.4 was used for obtaining a new substrate with good adsorption capacity for dyes and heavy metals from wastewater. A new material was obtained from modified fly ash with NaOH and hexadecyltrimethylammonium bromide (HTAB) a cationic surfactant. Contact time, optimum amount of substrate and the pH corresponding to 50 mL solution of pollutants were the parameters optimized for obtaining the maximum efficiency in the adsorption process. The optimized adsorption parameters were further used in thermodynamic and kinetic studies of the adsorption processes. The adsorption kinetic mechanisms, and the substrate capacities are further discussed correlated with the surface structure (XRD), composition (EDS, FTIR), and morphology (SEM, AFM). The results indicate that the novel nano-substrate composite with fly ash modified can be used as an efficient and low cost adsorbent for simultaneous removal of dyes and heavy metals, the resulted water respects the discharge regulations.

  20. Automatic power control for a pressurized water reactor

    International Nuclear Information System (INIS)

    During a normal operation of a pressurized water reactor (PWR), the reactivity is controlled by control rods, boron, and the average temperature of the primary coolant. Especially in load follow operation, the reactivity change is induced by changes in power level and effects of xenon concentration. The control of the core power distribution is concerned, mainly, with the axial power distribution which depends on insertion and withdrawal of the control rods resulting in additional reactivity compensation. The utilization of part strength control element assemblies (PSCEAs) is quite appropriate for a control of the power distribution in the case of Yonggwang Nuclear Unit 3 (YGN Unit 3). However, control of the PSCEAs is not automatic, and changes in the boron concentration by dilution/boration are done manually. Thus, manual control of the PSCEAs and the boron concentration require the operator's experience and knowledge for a successful load follow operation. In this thesis, the new concepts have been proposed to adapt for an automatic power control in a PWR. One of the new concepts is the mode K control, another is a fuzzy power control. The system in mode K control implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial power shape change, which allows automatic control of the axial power distribution. And the mode K enables precise regulation, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1000-MWe PWR which is a reference plant for the YGN Unit 3. The simulation results illustrate that the mode K would be

  1. Heavy Metal Concentrations in Maltese Potable Water

    Directory of Open Access Journals (Sweden)

    Roberta Bugeja

    2015-05-01

    Full Text Available This study evaluates the levels of aluminum (Al, cadmium (Cd, chromium (Cr, copper (Cu, iron (Fe, lead (Pb, nickel (Ni and zinc (Zn in tap water samples of forty localities from around the Maltese Islands together with their corresponding service supply reservoirs. The heavy metal concentrations obtained indicated that concentrations of the elements were generally below the maximum allowed concentration established by the Maltese legislation. In terms of the Maltese and EU water quality regulations, 17.5% of the localities sampled yielded water that failed the acceptance criteria for a single metal in drinking water. Higher concentrations of some metals were observed in samples obtained at the end of the distribution network, when compared to the concentrations at the source. The observed changes in metal concentrations between the localities’ samples and the corresponding supply reservoirs were significant. The higher metal concentrations obtained in the samples from the localities can be attributed to leaching in the distribution network.

  2. Heavy water in the context of hydrogen economy. Prospects for cheaper production by water electrolysis

    International Nuclear Information System (INIS)

    Hydrogen is an extremely important material. It is commonly used in many industrial processes. It can also be used as the key medium in 'hydrogen energy philosophy' due to its unique energetic properties (production for storage, gas-line transport). Its heavy isotopes, deuterium (D) and tritium (T), are very important nuclear materials. Deuterium, in the form of heavy water, is an excellent moderator in fission reactors, while both D and T are now seen as fuel components in fusion reactors in the future. Thus, improvements of production processes for hydrogen and its isotopes are always actual. Electrolysis (sometimes in combination with other methods) is often used for heavy water production or re-enrichment or for tritium removal from 'nuclear waters', mostly because of high D/H (T/H, T/D) isotope separation factors, although the electrolysis consumes great amounts of energy (about 4.5 to 5 kWh/m3 H2 in industrial electrolyzers). There were various attempts to improve this process: zero-gap cell geometry, development of new diaphragm materials, development of new electrocatalytic materials for electrodes, using so-called ionic activators etc. We investigated the use of catalytic cathode materials made from hypo-hyper-d-electronic combinations of transition metals as well as in situ activation of electrodes. Many intermetallic combinations were tried. Two types of ionic activators were used: tris-(ethylenediamine)-Co(III)-chloride complex and tris-(trimethylenediamine)-Co(III)-chloride complex. Some significant increases of the separation factors were obtained. Dependence of isotope enrichment on the amount of water that must be electrolysed for was estimated for different values of the separation factor. It was concluded that this a good way to increase the efficiency of the process by achieving an energy saving and an increase of the separation factors simultaneously. The method is discussed in a context that assumes heavy water as a by-product of the hydrogen

  3. Hydrogen and water reactor safety: proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  4. Mixed oxide fuel for water cooled reactors

    International Nuclear Information System (INIS)

    The problems connected with introduction of plutonium extracted from spent fuels of operating NPPs into water cooled reactor fuel cycle are considered. The trends in formation of the World market of mixed fuel are illustrated taking as examples Great Britain and Japan

  5. Hydrogen and water reactor safety: proceedings

    International Nuclear Information System (INIS)

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability

  6. Developmental Light-Water Reactor Program

    International Nuclear Information System (INIS)

    This report summarizes the progress of the Developmental Light-Water Reactor (DLWR) Program at Oak Ridge National Laboratory in FY 1989. It also includes (1) a brief description of the program, (2) definition of goals, (3) earlier achievements, and (4) proposed future activities

  7. Mitigation performance indicator for boiling water reactors

    International Nuclear Information System (INIS)

    All U.S. boiling water reactors (BWRs) inject hydrogen for mitigation of intergranular stress corrosion cracking (IGSCC), and most currently use or plan to use noble metals technology. The EPRI Boiling Water Reactor Vessels and Internals Project (BWRVIP) developed a Mitigation Performance Indicator (MPI) in 2006 to accurately depict to management the status of mitigation equipment and as a standardized way to show the overall health of reactor vessel internals from a chemistry perspective. It is a 'Needed' requirement in the EPRI BWR Water Chemistry Guidelines that plants have an MPI, and use of the BWRVIP MPI is a 'Good Practice'. The MPI is aligned with inspection relief criteria for reactor piping and internal components for U.S. BWRs. This paper discusses the history of the MPI, from its first use for plants operating with moderate hydrogen water chemistry (HWC-M) or Noble Metal Chemical Application (NMCA) + HWC to its more recent use for plants operating with On-Line NobleChem™ (OLNC) + HWC. Key mitigation parameters are discussed along with the technical bases for the indicators associated with the parameters. (author)

  8. Light-water reactor accident classification

    International Nuclear Information System (INIS)

    The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art

  9. AFRRI TRIGA Reactor water quality monitoring program

    International Nuclear Information System (INIS)

    AFRRI has started a water quality monitoring program to provide base line data for early detection of tank leaks. This program revealed problems with growth of algae and bacteria in the pool as a result of contamination with nitrogenous matter. Steps have been taken to reduce the nitrogen levels and to kill and remove algae and bacteria from the reactor pool. (author)

  10. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    The fuel assembly of a boiling water reactor contains a number of vertical fuel rods with their lower ends against a bottom tie plate. The rods are positioned by spacers, which are fixed to the canning. The upward motion is reduced by the top plate of a special design. (G.B.)

  11. Methane production by treating vinasses from hydrous ethanol using a modified UASB reactor

    Directory of Open Access Journals (Sweden)

    España-Gamboa Elda I

    2012-11-01

    Full Text Available Abstract Background A modified laboratory-scale upflow anaerobic sludge blanket (UASB reactor was used to obtain methane by treating hydrous ethanol vinasse. Vinasses or stillage are waste materials with high organic loads, and a complex composition resulting from the process of alcohol distillation. They must initially be treated with anaerobic processes due to their high organic loads. Vinasses can be considered multipurpose waste for energy recovery and once treated they can be used in agriculture without the risk of polluting soil, underground water or crops. In this sense, treatment of vinasse combines the elimination of organic waste with the formation of methane. Biogas is considered as a promising renewable energy source. The aim of this study was to determine the optimum organic loading rate for operating a modified UASB reactor to treat vinasse generated in the production of hydrous ethanol from sugar cane molasses. Results The study showed that chemical oxygen demand (COD removal efficiency was 69% at an optimum organic loading rate (OLR of 17.05 kg COD/m3-day, achieving a methane yield of 0.263 m3/kg CODadded and a biogas methane content of 84%. During this stage, effluent characterization presented lower values than the vinasse, except for potassium, sulfide and ammonia nitrogen. On the other hand, primers used to amplify the 16S-rDNA genes for the domains Archaea and Bacteria showed the presence of microorganisms which favor methane production at the optimum organic loading rate. Conclusions The modified UASB reactor proposed in this study provided a successful treatment of the vinasse obtained from hydrous ethanol production. Methanogen groups (Methanobacteriales and Methanosarcinales detected by PCR during operational optimum OLR of the modified UASB reactor, favored methane production.

  12. Fast reactors and advanced light water reactors for sustainable development

    International Nuclear Information System (INIS)

    Complete text of publication follows: The importance of nuclear energy, as a realistic option to solve the issues of the depletion of energy resources and the global environment, has been re-acknowledged worldwide. In response to this international movement, the papers compiling the most recent findings in the fields of fast reactors (FR) and advanced light water reactors (LWR) were gathered and published in this special issue. This special issue compiles six articles, most of which are very meticulously performed studies of the multi year development of design and assessment methods for large sodium-cooled FRs (SFRs), and two are related to the fuel cycle options that are leading to a greater understanding on the efficient utilization of energy resources. The Japanese sodium-cooled fast reactor (JSFR) is addressed in two manuscripts. H. Yamano et al. reviewed the current design which adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. Their safety assessments of both design basis accidents and severe accidents indicate that the devised JSFR satisfies well their risk target. T. Takeda et al. discussed the improvement of the modeling accuracy for the detailed calculation of JSFR's features in three areas: neutronics, fuel materials, and thermal hydraulics. The verification studies which partly use the measured data from the prototype FBR Monju are also described. Two of these manuscripts deal with those aspects of advanced design of SFR that have hitherto not been explored in great depth. The paper by G. Palmiotti et al. explored the possibility of using the sensitivity methodologies in the reactor physics field. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described. F. Baque et al. reviewed the evolution of the in

  13. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T.; Grunwald, G.

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  14. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  15. Transient following partial loss of feed water for thorium based natural circulation reactor

    International Nuclear Information System (INIS)

    The proposed Advanced Heavy Water Reactor (AHWR) is a 920 MWth vertical pressure tube type boiling light water cooled and heavy water moderated reactor. One of the important passive design features of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all allowed power levels with no primary coolant pumps. Apart from this passive design feature passive safety systems in AHWR include isolation condenser (IC) system for decay heat removal in case of unavailability of main steam condenser, emergency core cooling (includes both high pressure and low pressure ECC) system, Passive containment cooling system, Passive containment isolation and automatic depressurization system. Further, reactor core has negative void coefficient of reactivity at all power level which enhances the safety of the reactor. The Primary Heat Transport System of the reactor consists of reactor core, core inlet and outlet core bottom extensions, inlet feeders, tailpipes, steam drums, downcomer and inlet header. One BFP trip transient without standby pump initiation has been analysed. This partial loss of feed scenario leads to reactor trip and subsequent decay heat removal takes place through isolation condenser path. All other thermal hydraulic parameters remain within safety limits

  16. Uncommon water chemistry observations in modern day boiling water reactors

    International Nuclear Information System (INIS)

    Numerous technologies have been developed to mitigate intergranular stress corrosion cracking (IGSCC) of boiling water reactor (BWR) materials that include hydrogen water chemistry (HWC), noble metal chemical application (NMCA) and on-line NMCA (OLNC). These are matured technologies with extensive plant operating experiences, HWC – 32 years, NMCA – 18 years and OLNC – 9 years. Over the past three decades, numerous water chemistry data, dose rate data and IGSCC mitigation data relating to these technologies have been published and presented at many international conferences. However, there are many valuable and critical water chemistry and dose rate data that have gone unnoticed and unreported. The purpose of this paper is to highlight some of the uncommon water chemistry and dose rate experiences that reveal valuable information on the performance and durability of NMCA and OLNC technologies. Data will be presented, that have hitherto been unseen in public domain, from the lead OLNC plant in Switzerland giving reasons for some of the uncommon or overlooked water chemistry observations. They include, decreasing reactor water platinum concentration with each successive OLNC application, lack of increase in reactor water activation products in later applications, gradual disappearance of main steam line radiation (MSLR) monitor response decrease, Curium and Au-199 release during OLNC applications, rapid increase in reactor water clean-up conductivity, and Iodine, Mo-99 and Tc-99m spiking when hydrogen is interrupted and brought back to service, and main steam and reactor water conductivity spiking when clean-up beds or condensate demineralizers are changed. All these observations give valuable information on the success of OLNC applications and also signal the presence of sufficient noble metal on in-reactor surfaces from the long term durability and effectiveness stand point. Some of these observations can be used as secondary parameters, if and when a primary

  17. Modified clay sorbents for wastewater treatment and immobilization of heavy metals in soils

    Science.gov (United States)

    Burlakovs, Juris; Klavins, Maris; Vincevica-Gaile, Zane; Stapkevica, Mara

    2014-05-01

    Soil and groundwater pollution with heavy metals is the result of both, anthropogenic and natural processes in the environment. Anthropogenic influence in great extent appears from industry, mining, treatment of metal ores and waste incineration. Contamination of soil and water can be induced by diffuse sources such as applications of agrochemicals and fertilizers in agriculture, air pollution from industry and transport, and by point sources, e.g., wastewater streams, runoff from dump sites and factories. Treatment processes used for metal removal from polluted soil and water include methodologies based on chemical precipitation, ion exchange, carbon adsorption, membrane filtration, adsorption and co-precipitation. Optimal removal of heavy metal ions from aqueous medium can be achieved by adsorption process which is considered as one of the most effective methods due to its cost-effectiveness and high efficiency. Immobilization of metals in contaminated soil also can be done with different adsorbents as the in situ technology. Use of natural and modified clay can be developed as one of the solutions in immobilization of lead, zinc, copper and other elements in polluted sites. Within the present study clay samples of different geological genesis were modified with sodium and calcium chlorides, iron oxyhydroxides and ammonium dihydrogen phosphate in variable proportions of Ca/P equimolar ratio to test and compare immobilization efficiency of metals by sorption and batch leaching tests. Sorption capacity for raw clay samples was considered as relatively lower referring to the modified species of the same clay type. In addition, clay samples were tested for powder X-ray difractometry, cation exchange, surface area properties, elemental composition, as well as scanning electron microscopy pictures of clay sample surface structures were obtained. Modified clay sorbents were tested for sorption of lead as monocontaminant and for complex contamination of heavy metals. The

  18. Thermophysical properties of saturated light and heavy water for advanced neutron source applications

    Energy Technology Data Exchange (ETDEWEB)

    Crabtree, A.; Siman-Tov, M.

    1993-05-01

    The Advanced Neutron Source is an experimental facility being developed by Oak Ridge National Laboratory. As a new nuclear fission research reactor of unprecedented flux, the Advanced Neutron Source Reactor will provide the most intense steady-state beams of neutrons in the world. The high heat fluxes generated in the reactor [303 MW(t) with an average power density of 4.5 MW/L] will be accommodated by a flow of heavy water through the core at high velocities. In support of this experimental and analytical effort, a reliable, highly accurate, and uniform source of thermodynamic and transport property correlations for saturated light and heavy water were developed. In order to attain high accuracy in the correlations, the range of these correlations was limited to the proposed Advanced Neutron Source Reactor's nominal operating conditions. The temperature and corresponding saturation pressure ranges used for light water were 20--300[degrees]C and 0.0025--8.5 MPa, respectively, while those for heavy water were 50--250[degrees]C and 0.012--3.9 MPa. Deviations between the correlation predictions and data from the various sources did not exceed 1.0%. Light water vapor density was the only exception, with an error of 1.76%. The physical property package consists of analytical correlations, SAS codes, and FORTRAN subroutines incorporating these correlations, as well as an interactive, easy-to-use program entitled QuikProp.

  19. Penn State advanced light water reactor concept

    International Nuclear Information System (INIS)

    The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

  20. Electrochemistry of Water-Cooled Nuclear Reactors

    International Nuclear Information System (INIS)

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or ''radiation fields'' around the primary loop and the vessel, as a function of the operating parameters and the water chemistry

  1. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  2. Towards intrinsically safe light-water reactors

    International Nuclear Information System (INIS)

    Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification

  3. Removal of heavy metals from water by zeolite mineral chemically modified. Mercury as a particular case; Remocion de metales pesados del agua por mineral zeolitico quimicamente modificado. Mercurio como un caso particular

    Energy Technology Data Exchange (ETDEWEB)

    Gebremedhin H, T

    2002-07-01

    Research works on the removal of mercury from water by zeolite minerals show that a small quantity of this element is sorbed. In this work the mercury sorption from aqueous solutions in the presence and absence of Cu(l l), Ni(l l) and/or Zn(l l) by a Mexican zeolite mineral, natural and modified by cisteaminium chloride or cistaminium dichloride, was investigated in acidic p H. The zeolite minerals were characterized by X- Ray diffraction Ftir, scanning electron microscopy and semiquantitative elemental analysis (EDS), surface area analysis (BET) and thermogravimetric analysis (TGA). Mercury from aqueous solutions was quantified by Atomic absorption spectroscopy. The amount of sulphur on the zeolite samples treated with Na CI and modified with cisteaminium chloride (0.375 mmol/g) or cistaminium dichloride(0.475 mmol/g) was found to be higher than that of the zeolite minerals modified with cisteaminium chloride and cistaminium dichloride without treating them with Na CI. The amount of sulphur on the zeolite minerals modified with thiourea was the lowest. The diffusion coefficients and sorption isotherms for mercury were determined in the natural, treated with Na CI and, treated with Na CI and then modified with the cisteaminium chloride or cistaminium dichloride zeolite samples. The retention of mercury was the highest for the zeolite minerals treated Na CI and then modified with cisteaminium chloride or cistaminium dichloride, with adsorption capacity of 0.0511 and 0.0525 mmol Hg/g, respectively. In this research work, it was found that the retention of mercury by the modified minerals was not affected by the presence of Cu (Il), Zn(l l) y Ni (I l) under the experimental conditions. (Author)

  4. Behavior of water reactor fuel rod

    International Nuclear Information System (INIS)

    This paper reviewed the fuels used widely in forms of (1) Zircaloy-sheathed UO2 fuel in light water-commercial power reactor, (2) Zircaloy-sheathed PuO2-UO2 fuel in plutonium-thermal reactor and advanced reactor (ATR), (3) aluminide and silicide fuel in Material Testing Reactors. From fundamental view points, physical/chemical properties and irradiation behaviors of both fuels and zircaloy claddings are briefly reviewed in chapters 1 and 2. Change of the fuel rod physical parameters with progress of burn-up are summed up in chapter 3. Some fuel troubles and failures encountered in past usage of worldwide LWR fuels are introduced with counterplans taken. In the last session of this chapter, recent results of R and D works have been carried out by fuel vendors are reviewed. Especially, in-core behaviors of PCI-remedy fuels developed to use for high burn-up extension and for load-follow operation are highlighted. Reactor accidents occurred through past forty years are surveyed and reviewed. Fuel behaviors during the reactivity initiated accident (RIA), the power-coolant mismatch (PCM), and the loss-of-coolant accident (LOCA) are taken into this review by using disclosed literatures. Safety criteria being used in Japanese licensing authorities are introduced relating to the fuel design limit. (author)

  5. Advanced light water reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    Giedraityte, Zivile [Helsinki University of Technology, Otaranta 8D-84, 02150 Espoo (Finland)

    2008-07-01

    For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to 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 which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

  6. 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.

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

    International Nuclear Information System (INIS)

    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.

  8. Guidebook on non-destructive examination of water reactor fuel

    International Nuclear Information System (INIS)

    To date, a significant quantity of data has been collected and published on power reactor fuel examination to determine the performance when subjected to radiation. The data have been published in technical reports and papers in technical journals. However, the usefulness of the published data to the IAEA Member States is limited. This is due to a number of reasons, including the large variety of examination methods, incomplete documentation of the data and lack of sufficiently detailed information on pre-irradiation data and irradiation history. To alleviate some of these problems, the Agency initiated a Co-ordinated Research Programme in 1983 entitled ''Examination and Documentation Methodology for Water Reactor Fuel''. The programme meetings usually involved technical contributions from the programme participants, followed by a detailed discussion of the various examination methods presented in these contributions. Based on these discussions and contributions, a guidebook on the examination and documentation methodology for light and heavy water reactor fuel has been prepared. The guidebook addresses the most commonly used examination methods for the various water reactor fuel systems. Limitations of each of the measurement techniques are also discussed, including their accuracy and precision. A detailed description of the measurement equipment is given and the common methods of documenting the data are also addressed. With the adoption of the uniform set of procedures and documentation methods, it is hoped that the IAEA Member States will be able to use effectively both the existing data and the future data from the various national programmes. It is also expected that this guidebook will be useful for adaptation of measurement techniques that are unique to specific fuel systems to other fuel types. 59 refs, 33 figs, 4 tabs

  9. Heavy ion irradiation of crystalline water ice

    CERN Document Server

    Dartois, E; Boduch, P; Brunetto, R; Chabot, M; Domaracka, A; Ding, J J; Kamalou, O; Lv, X Y; Rothard, H; da Silveira, E F; Thomas, J C

    2015-01-01

    Under cosmic irradiation, the interstellar water ice mantles evolve towards a compact amorphous state. Crystalline ice amorphisation was previously monitored mainly in the keV to hundreds of keV ion energies. We experimentally investigate heavy ion irradiation amorphisation of crystalline ice, at high energies closer to true cosmic rays, and explore the water-ice sputtering yield. We irradiated thin crystalline ice films with MeV to GeV swift ion beams, produced at the GANIL accelerator. The ice infrared spectral evolution as a function of fluence is monitored with in-situ infrared spectroscopy (induced amorphisation of the initial crystalline state into a compact amorphous phase). The crystalline ice amorphisation cross-section is measured in the high electronic stopping-power range for different temperatures. At large fluence, the ice sputtering is measured on the infrared spectra, and the fitted sputtering-yield dependence, combined with previous measurements, is quadratic over three decades of electronic ...

  10. Neutronic calculations in heavy water moderated multiplying media using GGC-3 library nuclear data

    International Nuclear Information System (INIS)

    Differences in obtaining transference matrices between GGC-3 code and the system to produce multigroup cross sections using GGC-3 library, recently implemented at the Neutrons and Reactors Division, have been analized. Neutronic calculations in multiplicative systems containing heavy water have been made using both methods. From the obtained results, it is concluded that the new method is more appropriate to deal with systems including moderators other than light water. (author)

  11. An analysis of reactor transient response for boiling water reactor ATWS events

    International Nuclear Information System (INIS)

    Numerical simulations of BWR (boiling water reactor) dynamic response under ATWS (anticipated transient without scram) conditions are presented for the case where the reactor is operated at natural circulation conditions. In non-isolation events, reactor stability is strongly influenced by the degree of core inlet subcooling. At normal water level and pressure, instabilities develop if core-inlet subcooling exceeds a critical value of ∼21 Btu/lbm. A sensitivity study with regard to the steam separator pressure - loss coefficient, however, indicates that system stability is strongly dependent on the magnitude of this parameter which suggests a significant degree of uncertainty In the results. Under isolation conditions at rated pressure, stability is significantly enhanced by rapid pressure fluctuations generated through cycling of safety/relief valves. Large-amplitude instabilities develop, however, in depressurization events, and SRV cycling no longer stabilizes the system. In a simulated depressurization to ∼500 psia, prompt critical excursions occurred, and oscillation amplitudes reached 1000% of rated power. Implications of the Present Study: With the exception of guidance to avoid SRV cycling, these preliminary results have provided further support for the validity of this response strategy. SABRE calculations have shown that the reactor is probably slightly unstable in natural circulation operation, but relief valve cycling prevents the occurrence of instabilities at or near design pressure. Thus, unstable operation should not be a concern when boron injection and HPCI are available and depressurization is unnecessary. The reactor water level for injection flow corresponding to HPCI operation has been shown to be acceptable and consistent with earlier estimates based on the NSAC results; however, condensation effects have a significant influence on the equilibrium reactor water level. In addition, the SABRE results reinforce PP and L's concerns that

  12. Liquid-cooled nuclear reactor, especially a boiling water reactor

    International Nuclear Information System (INIS)

    A nuclear reactor with a special arrangement of fuel rods in the core is designed. Each fuel element has its shaft which is made of sheets, has the same cross section as the fuel element and protrudes at least the length of the control rod above the reactor core. Made of a zirconium alloy in the core area and of stainless steel above it, the shaft is equipped with channels for sliding the rods in and out and serves to spatially secure the position of the rods. Coolant flow is provided by the chimney effect. The shaft can conveniently enclose the control rod drive. It can also serve to bear the water separator. Moreover, it can constitute a part of the casing which surrounds the fuel rods and keeps the fuel in an intimate contact with the coolant; the other part of this casing is constituted by inserted sheets which can conveniently have the shape of angles. The walls of neighboring shafts form a compartment accommodating a neutron absorber plate. (M.D.). 11 figs

  13. Microflora of nuclear research reactor pool water

    International Nuclear Information System (INIS)

    The circulation of pool water through the nuclear reactor core produces a bactericidal effect on the microflora due to the influence of various kinds of radiation. The microbe contents return to their initial level in 2 to 4 months after the circulation has stopped. The microflora comprises mainly cocci in large numbers, G-positive rods and fungi, and lower amounts of G-negative rods as compared with the water with which the reactor pool was initially filled. Increased amounts are present of radiation-resistant forms exhibiting intense production of catalase and nuclease. Supposedly, the presence of these enzymes is in some way beneficial to the microbes in their survival in the high-radiation zones. (author). 1 fig., 2 tabs., 12 refs

  14. Fission gas behaviour in water reactor fuels

    International Nuclear Information System (INIS)

    During irradiation, nuclear fuel changes volume, primarily through swelling. This swelling is caused by the fission products and in particular by the volatile ones such as krypton and xenon, called fission gas. Fission gas behaviour needs to be reliably predicted in order to make better use of nuclear fuel, a factor which can help to achieve the economic competitiveness required by today's markets. These proceedings communicate the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors. (author)

  15. Resolution of US regulatory issues involving boiling water reactor stability

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission (NRC) and the Boiling Water Reactor Owners Group (BWROG) have been reexamining BWR instability characteristics and consequences since the March 1988 instability event at LaSalle Unit 2. The NRC and BWROG concluded that existing reactor protection systems do not prevent violation of the critical power ratio (CPR) safety limits caused by large asymmetric oscillations. The studies are also examining the need to modify the automatic and operator actions previously developed for response to an anticipated transient without scram (ATWS) event because of oscillation effects not fully considered in previous studies. This paper presents the current status of these studies and an assessment of actions needed to resolve the issue. (author)

  16. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

    A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

  17. Consequence of potential accidents in heavy water plants

    International Nuclear Information System (INIS)

    Heavy water plants realize the primary isotopic concentrations of water using H2O-H2S chemical exchange and they are chemical plants. As these plants are handling and spreading large quantities of hydrogen sulphide (high toxic, corrosive, flammable and explosive as) maintained in the process at relative high temperatures and pressures, it is required an assessing of risks associated with the potential accidents. The H2S released in atmosphere as a result of an accident will have negative consequences to property, population and environment. This paper presents a model of consequences quantitative assessment and its outcome for the most dangerous accident in heavy water plants. Several states of the art risk based methods were modified and linked together to form a proper model for this analyse. Five basic steps to identify the risks involved in operating the plants are followed: hazard identification, accident sequence development, H2S emissions calculus, dispersion analyses and consequences determination. A brief description of each step and some information of analysis results are provided. The accident proportions, the atmospheric conditions and the population density in the respective area were accounted for consequences calculus. The specific results of the consequences analysis allow to develop the plant's operating safety requirements so that the risk remain at an acceptable level. (authors)

  18. Environmentally assisted cracking in light water reactors. Semiannual report, April--September 1991: Volume 13

    Energy Technology Data Exchange (ETDEWEB)

    Kassner, T F; Ruther, W E; Chung, H M; Hicks, P D; Hins, A G; Park, J Y; Soppet, W K; Shack, W J [Argonne National Lab., IL (United States)

    1992-03-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with {approx} 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289{degrees}C.

  19. Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Chopra, O.K.; Erck, R.A.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

    1995-03-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials.

  20. Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

    International Nuclear Information System (INIS)

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289 degrees C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials

  1. Water chemistry of breeder reactor steam generators

    International Nuclear Information System (INIS)

    The water quality requirements will be described for breeder reactor steam generators, as well as specifications for balance of plant protection. Water chemistry details will be discussed for the following power plant conditions: feedwater and recirculation water at above and below 5% plant power, refueling or standby, makeup water, and wet layup. Experimental data will be presented from tests which included a departure from nucleate boiling experiment, the Few Tube Test, with a seven tube evaporator and three tube superheater, and a verification of control and on-line measurement of sodium ion in the ppB range. Sampling and instrumentation requirements to insure adherence to the specified water quality will be described. Evaporator cleaning criteria and data from laboratory testing of chemical cleaning solutions with emphasis on flow, chemical composition, and temperature will be discussed

  2. Burnup determination of water reactor fuel

    International Nuclear Information System (INIS)

    The present meeting was scheduled by the International Atomic Energy Agency in consultation with the Members of the International Working Group on Water Reactor Fuel Performance and Technology. The meeting was hosted by the Commission of the European Communities, at the Transuranium Research Laboratory, Joint Research Centre Karlsruhe, in the Federal Republic of Germany. This subject was dealt with for the first time by the IAEA. It was found to correspond adequately to this type of Specialist Meeting and to be suitable at a moment when the extension of burnup constitutes a major technical and economical issue in fuel technology. It was stressed that analysis of highly burnt fuels, mixed oxides and burnable absorber bearing fuels required extension of the experimental data base, to comply with the increasing demand for an improved fuel management, including better qualification of reactor physics codes. Twenty-seven participants from eleven countries plus two international organizations attended the Meeting. Twelve papers were given during three technical sessions, followed by a panel discussion which allowed to formulate the conclusions of the meeting and recommendations to the Agency. In addition, participants were invited to give an outline of their national programmes, related to Burnup Determination of Water Reactor Fuel. A separate abstract was prepared for each of these 12 papers. Refs, figs and tabs

  3. The decommissioning of the water boiler reactor

    International Nuclear Information System (INIS)

    Following completion of service, the Water Boiler Reactor (WBR) has been decommissioned by the Institute of Nuclear Energy Research (INER) under the Atomic Energy Council's (AEC) regulation. The WBR is a light water moderated and graphite reflected research reactor with peak thermal power of 100 kW. The unique feature of the WBR is that it is fueled with uranyl sulfate (UO2SO4) which is in liquid form. Since there is another research reactor owned by I7NER of megawatt scale in the planning stages for decommissioning, the WBR project was conducted with great care to accumulate experience. Extensive planning by INER and step-by-step regulative activities by AEC were followed regardless of the structural simplicity of the WBR. Valuable information was gathered in the task and will be useful for preparing future decommissioning needs. The major work in the WBR decommissioning project was finished within six months and the accumulated dose received during the work was 1 9.63mSv. (author)

  4. Dynamic modelling of Industrial Heavy Water Plant

    International Nuclear Information System (INIS)

    The dynamic behavior of the isotopic enrichment unites of the Industrial Heavy Water Plant, located in Arroyito, Neuquen, Argentina, was modeled and simulated in the present work. Dynamic models of the chemical and isotopic interchange processes existent in the plant, were developed. This served as a base to obtain representative models of the different unit and control systems. The developed models were represented in a modular code for each unit. Each simulator consists of approximately one hundred non-linear-first-order differential equations and some other algebraic equation, which are time resolved by the code. The different simulators allow to change a big number of boundary conditions and the control systems set point for each simulation, so that the program become very versatile. The output of the code allows to see the evolution through time of the variables of interest. An interface which facilitates the use of the first enrichment stage simulator was developed. This interface allows an easy access to generate wished events during the simulation and includes the possibility to plot evolution of the variables involved. The obtained results agree with the expected tendencies. The calculated nominal steady state matches by the manufacturer. The different steady states obtained, agree with previous works. The times and tendencies involved in the transients generated by the program, are in good agreement with the experience obtained at the plant. Based in the obtained results, it is concluded that the characteristic times of the plant are determined by the masses involved in the process. Different characteristics in the system dynamic behavior were generated with the different simulators, and were validated by plant personnel. This work allowed to understand the different process involved in the heavy water manufacture, and to develop a very useful tool for the personnel of the plant. (author). 14 refs., figs., tabs. plant. (author). 14 refs., figs., tabs

  5. Assessment of released heavy metals from electrical and electronic equipment (EEE) existing in shipwrecks through laboratory-scale simulation reactor

    International Nuclear Information System (INIS)

    Highlights: ► A laboratory-scale reactor was built to simulate the “Sea Diamond” shipwreck. ► EEE was thrown into the reactor for heavy metal release rate assessment. ► 15 seawater samples were taken and analyzed in a nine month experimental period. ► Zinc, mercury and copper were found in concentrations above the CMC criterion. ► Nickel and lead were found in concentrations higher than the CCC criterion. -- Abstract: In a passenger ship, the existence of EEE is obvious. In time, under shipwreck's conditions, all these materials will undergo an accelerated severe corrosion, due to salt water, releasing, consequently, heavy metals and other hazardous substances in the aquatic environment. In this study, a laboratory-scale reactor was manufactured in order to simulate the conditions under which the “Sea Diamond” shipwreck lies (14 bars of pressure and 16 °C of temperature) and remotely observe and assess any heavy metal release that would occur, from part of the EEE present in the ship, into the sea. Ten metals were examined and the results showed that zinc, mercury and copper were abundant in the water samples taken from the reactor and in significantly higher concentrations compared to the US EPA CMC (criterion maximum concentration) criterion. Moreover, nickel and lead were found in concentrations higher than the CCC (criterion constant concentration) criterion set by the US EPA for clean seawater. The rest of the elements were measured in concentrations within the permissible limits. It is therefore of environmental benefit to salvage the wreck and recycle all the WEEE found in it

  6. Controlling hydrogen behavior in light water reactors

    International Nuclear Information System (INIS)

    In the aftermath of the incident at Three Mile Island Unit 2 (TMI-2), a new and different treatment of the Light Water Reactor (LWR) risks is needed for public safety because of the specific events involving hydrogen generation, transport, and behavior following the core damage. Hydrogen behavior in closed environments such as the TMI-2 containment building is a complex phenomenon that is not fully understood. Hence, an engineering approach is presented for prevention of loss of life, equipment, and environment in case of a large hydrogen generation in an LWR. A six-level defense strategy is described that minimizes the possibility of ignition of released hydrogen gas and otherwise mitigates the consequences of hydrogen release. Guidance is given to reactor manufacturers, utility companies, regulatory agencies, and research organizations committed to reducing risk factors and insuring safety of life, equipment, and environment

  7. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  8. Mathematical modelling of water radiolysis kinetics under reactor conditions

    International Nuclear Information System (INIS)

    Experimental data on coolant radiolysis (RBMK-1000 reactor) were used to construct mathematical model of water radiolysis kinetics under reactor conditions. Good agreement of calculation results with the experiment is noted

  9. Environmentally assisted cracking in Light Water Reactors: Semiannual report, October 1994--March 1995. Volume 20

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Chopra, O.K.; Gavenda, D.J.; Hins, A.G.; Kassner, T.F.; Ruther, W.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

    1996-01-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several dissolvedoxygen (DO) concentrations to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Tensile properties and microstructures of several heats of Alloy 600 and 690 were characterized for correlation with EAC of the alloys in simulated LWR environments. Effects of DO and electrochemical potential on susceptibility to intergranular cracking of high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath irradiated in boiling water reactors were determined in slow-strain-rate-tensile tests at 289{degrees}C. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  10. Experimental and analytical study on thermal hydraulics in reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Akimoto, Hajime; Araya, Fumimasa; Ohnuki, Akira; Yoshida, Hiroyuki; Kureta, Masatoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-06-01

    Study and development of reduced-moderation spectrum water reactor proceeds as a option of the future type reactor in Japan Atomic Energy Research Institute (JAERI). The reduced-moderation spectrum in which a neutron has higher energy than the conventional water reactors is achieved by decreasing moderator-to-fuel ratio in the lattice core of the reactor. Conversion ratio in the reduced-moderation water reactor can be more than 1.0. High burnup and long term cycle operation of the reactor are expected. A type of heavy water cooled PWR and three types of BWR are discussed as follows; For the PWR, (1) critical heat flux experiments in hexagonal tight lattice core, (2) evaluation of cooling limit at a nominal power operation, and (3) analysis of rewetting cooling behavior at loss of coolant accident following with large scale pipe rupture. For the BWR, analyses of cooling limit at a nominal power operation of, (1) no blanket BWR, (2) long term cycle operation BWR, and (3) high conversion ratio BWR. The experiments and the analyses proved that the basic thermal hydraulic characteristics of these reduced-moderation water reactors satisfy the essential points of the safety requirements. (Suetake, M.)

  11. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    International Nuclear Information System (INIS)

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point

  12. RELAP5-3D code for supercritical-pressure, light-water-cooled reactors

    International Nuclear Information System (INIS)

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point. (author)

  13. Environmentally assisted cracking in light-water reactors: Semi-annual report, January--June 1997. Volume 24

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E. [Argonne National Lab., IL (United States)] [and others

    1998-04-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1997 to June 1997. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Types 304 and 304L SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle is equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in low-DO, simulated pressurized water reactor environments.

  14. Silver-modified clinoptilolite for the removal of Escherichia coli and heavy metals from aqueous solutions.

    Science.gov (United States)

    Akhigbe, Lulu; Ouki, Sabeha; Saroj, Devendra; Lim, Xiang Min

    2014-09-01

    This paper investigates the potential of using the silver antibacterial properties combined with the metal ion exchange characteristics of silver-modified clinoptilolite to produce a treatment system capable of removing both contaminants from aqueous streams. The results have shown that silver-modified clinoptilolite is capable of completely eliminating Escherichia coli after 30-min contact time demonstrating its effectiveness as a disinfectant. Systems containing both E. coli and metals exhibited 100 % E. coli reduction after 15-min contact time and maximum metal adsorption removal efficiencies of 97, 98, and 99 % for Pb(2+), Cd(2+), and Zn(2+) respectively after 60 min; 0.182-0.266 mg/g of metal ions were adsorbed by the zeolites in the single- and mixed-metal-containing solutions. Nonmodified clinoptilolite showed no antibacterial properties. This study demonstrated that silver-modified clinoptilolite exhibited high disinfection and heavy metal removal efficiencies and consequently could provide an effective combined treatment system for the removal of E. coli and metals from contaminated water streams.

  15. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  16. Is light water reactor technology sustainable?

    International Nuclear Information System (INIS)

    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)

  17. Simulation of Boiling Water Reactor dynamics

    International Nuclear Information System (INIS)

    This master thesis describes a mathematical model of a boiling water reactor and address the dynamic behaviour of the neutron kinetics, boilding dynamics and pressur stability. The simulation have been done using the SIMNON-program. The meaning were that the result from this work possibly would be adjust to supervision methods suitable for application in computer systems. This master thesis in automatic control has been done at the Department of Automatic Control, Lund Institute of Technology. The initiative to the work came from Sydkraft AB. (author)

  18. 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)

  19. Method and apparatus for enrichment or upgrading heavy water

    International Nuclear Information System (INIS)

    A method and apparatus for upgrading and final enrichment of heavy water are described, comprising means for contacting partially enriched heavy water feed in a catalyst column with hydrogen gas (essentially D2) originating in an electrolysis cell so as to enrich the feed water with deuterium extracted from the electrolytic hydrogen gas and means for passing the deuterium enriched water to the electrolysis cell. (author)

  20. Emergency reactor core cooling water injection device for light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oda, Junro.

    1994-05-13

    A reactor pressure vessel is immersed in pool water of a reactor container. A control valve is interposed to a water supplying pipelines connecting pool water and a pressure vessel. A valve actuation means for opening/closing the control valve comprises a lifting tank. The inner side of the lifting tank and the inner side of the pressure vessel are connected by a communication pipeline (a syphon pipe) at upper and lower two portions. The lifting tank and the control valve are connected by a link mechanism. When a water level in the pressure vessel is lowered, the water level in the lifting tank is lowered to the same level as that in the pressure vessel. This reduces the weight of the lifting tank, the lifting tank is raised, to open the control valve by way of a link mechanism. As a result, liquid phase in the pressure vessel is in communication with the pool water, and the pool water flows down into the pressure vessel to maintain the reactor core in a flooded state. (I.N.).

  1. Performance of the heavy fraction of pyrolysis oil derived from waste printed circuit boards in modifying asphalt.

    Science.gov (United States)

    Yang, Fan; Sun, Shuiyu; Zhong, Sheng; Li, Shenyong; Wang, Yi; Wu, Jiaqi

    2013-09-15

    The focus of this research was the development of efficient and affordable asphalt modifiers. Pyrolysis oil was produced as a byproduct from the pyrolysis of waste printed circuit boards (WPCBs). The high boiling point fraction was separated from the pyrolysis oil through distillation and is referred to as the heavy fraction of pyrolysis oil (HFPO). The HFPO was tested as an asphalt modifier. Three asphalt modifiers were tested: HFPO; styrene-butadiene rubber (SBR); and HFPO + SBR (1:1). The physical properties and road performance of the three modified asphalts were measured and evaluated. The results have shown that when the amount of modifier was less than 10%, the HFPO modified asphalt had the highest softening point of the three. The dynamic stability (DS) and water resistance of the asphalt mixture with the HFPO modified asphalt was 10,161 cycles/mm and 87.2%, respectively. The DS was much larger than for the HFPO + SBR and SBR modified asphalt mixtures. These results indicate that using HFPO as an asphalt modifier has significant benefits not only for road engineering but also for resource recycling. PMID:23644664

  2. Pressurized water-reactor feedwater piping response to water hammer

    International Nuclear Information System (INIS)

    The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

  3. Assessment of irradiation effects on beryllium reflector and heavy water tank of JRR-3M

    Energy Technology Data Exchange (ETDEWEB)

    Murayama, Yoji; Kakehuda, Kazuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    The JRR-3M, a swimming pool type research reactor with beryllium and heavy water reflectors, has been operated since 1990. Since the beryllium reflectors are close to fuel and receive high fast neutron fluence in a relatively short time, they may be subject to change their dimensions by swelling due mostly to entrapped helium gaseous. This may bend the reflectors to the outside and narrow gaps between the reflectors and the fuel elements. The gaps have been measured with an ultrasonic thickness gage in an annual inspection. The results in 1996 show that the maximum of expansion in the diametral directions was 0.6 mm against 1.6 mm of a managed value for replacement of the reflector. A heavy water tank of the JRR-3M is made of aluminum alloy A5052. Surveillance tests of the alloy have been conducted to evaluate irradiation effects of the heavy water tank. Five sets of specimens of the alloy have been irradiated in the beryllium reflectors where fast neutron flux is higher than that in the heavy water tank. In 1994, one set of specimens had been unloaded and carried out the post-irradiation tests. The results show that the heavy water tank preserved satisfactory mechanical properties. (author)

  4. Self-Sustaining Thorium Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

    Full Text Available A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

  5. Storage of water reactor spent fuel in water pools. Survey of world experience

    International Nuclear Information System (INIS)

    Following discharge from a nuclear reactor, spent fuel has to be stored in water pools at the reactor site to allow for radioactive decay and cooling. After this initial storage period, the future treatment of spent fuel depends on the fuel cycle concept chosen. Spent fuel can either be treated by chemical processing or conditioning for final disposal at the relevant fuel cycle facilities, or be held in interim storage - at the reactor site or at a central storage facility. Recent forecasts predict that, by the year 2000, more than 150,000 tonnes of heavy metal from spent LWR fuel will have been accumulated. Because of postponed commitments regarding spent fuel treatment, a significant amount of spent fuel will still be held in storage at that time. Although very positive experience with wet storage has been gained over the past 40 years, making wet storage a proven technology, it appears desirable to summarize all available data for the benefit of designers, storage pool operators, licensing agenices and the general public. Such data will be essential for assessing the viability of extended water pool storage of spent nuclear fuel. In 1979, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD jointly issued a questionnaire dealing with all aspects of water pool storage. This report summarizes the information received from storage pool operators

  6. Zeolite Membrane Reactor for Water Gas Shift Reaction for Hydrogen Production

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Jerry Y.S. [Arizona State Univ., Mesa, AZ (United States)

    2013-01-29

    Gasification of biomass or heavy feedstock to produce hydrogen fuel gas using current technology is costly and energy-intensive. The technology includes water gas shift reaction in two or more reactor stages with inter-cooling to maximize conversion for a given catalyst volume. This project is focused on developing a membrane reactor for efficient conversion of water gas shift reaction to produce a hydrogen stream as a fuel and a carbon dioxide stream suitable for sequestration. The project was focused on synthesizing stable, hydrogen perm-selective MFI zeolite membranes for high temperature hydrogen separation; fabricating tubular MFI zeolite membrane reactor and stable water gas shift catalyst for membrane reactor applications, and identifying experimental conditions for water gas shift reaction in the zeolite membrane reactor that will produce a high purity hydrogen stream. The project has improved understanding of zeolite membrane synthesis, high temperature gas diffusion and separation mechanisms for zeolite membranes, synthesis and properties of sulfur resistant catalysts, fabrication and structure optimization of membrane supports, and fundamentals of coupling reaction with separation in zeolite membrane reactor for water gas shift reaction. Through the fundamental study, the research teams have developed MFI zeolite membranes with good perm-selectivity for hydrogen over carbon dioxide, carbon monoxide and water vapor, and high stability for operation in syngas mixture containing 500 part per million hydrogen sulfide at high temperatures around 500°C. The research teams also developed a sulfur resistant catalyst for water gas shift reaction. Modeling and experimental studies on the zeolite membrane reactor for water gas shift reaction have demonstrated the effective use of the zeolite membrane reactor for production of high purity hydrogen stream.

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

    International Nuclear Information System (INIS)

    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

  8. An optical dosimeter for monitoring heavy metal ions in water

    Science.gov (United States)

    Mignani, Anna G.; Regan, Fiona; Leamy, D.; Mencaglia, A. A.; Ciaccheri, L.

    2005-05-01

    This work presents an optochemical dosimeter for determining and discriminating nickel, copper, and cobalt ions in water that can be used as an early warning system for water pollution. An inexpensive fiber optic spectrophotometer monitors the sensor's spectral behavior under exposure to water solutions of heavy metal ions in the 1-10 mg/l concentration range. The Principal Component Analysis (PCA) method quantitatively determines the heavy metals and discriminates their type and combination.

  9. Screening reactor steam/water piping systems for water hammer

    International Nuclear Information System (INIS)

    A steam/water system possessing a certain combination of thermal, hydraulic and operational states, can, in certain geometries, lead to a steam bubble collapse induced water hammer. These states, operations, and geometries are identified. A procedure that can be used for identifying whether an unbuilt reactor system is prone to water hammer is proposed. For the most common water hammer, steam bubble collapse induced water hammer, six conditions must be met in order for one to occur. These are: (1) the pipe must be almost horizontal; (2) the subcooling must be greater than 20 C; (3) the L/D must be greater than 24; (4) the velocity must be low enough so that the pipe does not run full, i.e., the Froude number must be less than one; (5) there should be void nearby; (6) the pressure must be high enough so that significant damage occurs, that is the pressure should be above 10 atmospheres. Recommendations on how to avoid this kind of water hammer in both the design and the operation of the reactor system are made

  10. International Working Group on Fast Reactors meeting of specialists on sodium-water reactions. Summary report

    International Nuclear Information System (INIS)

    The meeting of specialists on sodium-water reactions was the first of its kind within the framework of activity of the International Working Group on Fast Reactors. The meeting was held at Argonne in accordance with the recommendation of the IWGPR and in agreement with the USA authorities. Participants from five countries took part in the meeting. The Agenda includes: Fundamental studies of reaction kinetics; Instrumentation for rig use; Sodium-water reaction tests and evaluation of results: - Large leaks; - Small leaks; Small-scale tests of corrosion/ erosion, metal wastage, etc. in presence of reacting sodium and water; Origin of water leak failure and its protection; Methods for detecting steam generator leaks and development of instrumentation for detection of sodium-water reactions in an actual steam generator; Test results and conclusions from test programs applicable to steam generator designs; Computing codes for the effect of a sodium-water reaction in a steam generating plant; Comparison with experiment; Influence of Geometry; Design of steam generators to accommodate reactions: - Methods for calculating effects of reaction on steam generator structure and secondary system; - Description of particular steam generators and associated sodium-water reaction effluent system, including ancillary circuits, instrumentation, etc.; - Design philosophy; - Experience of sodium-water reactions in steam generators; Design and effectiveness of relief systems; Methods for system recovery following a steam generator leak; Proposed further tests; Simulation tests of reactions using other fluids

  11. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  12. State space modeling of reactor core in a pressurized water reactor

    Science.gov (United States)

    Ashaari, A.; Ahmad, T.; Shamsuddin, Mustaffa; M, Wan Munirah W.; Abdullah, M. Adib

    2014-07-01

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  13. Environmentally assisted cracking in light water reactors annual report January - December 2005.

    Energy Technology Data Exchange (ETDEWEB)

    Alexandreanu, B.; Chen, Y.; Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.

    2007-08-31

    This report summarizes work performed from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs). Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed. The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45-3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed. The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa. Crack growth rate data

  14. Commercial Light Water Reactor Tritium Extraction Facility

    Energy Technology Data Exchange (ETDEWEB)

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  15. Radiation Protection at Light Water Reactors

    CERN Document Server

    Prince, Robert

    2012-01-01

    This book is aimed at Health Physicists wishing to gain a better understanding of the principles and practices associated with a light water reactor (LWR) radiation protection program. The role of key program elements is presented in sufficient detail to assist practicing radiation protection professionals in improving and strengthening their current program. Details related to daily operation and discipline areas vital to maintaining an effective LWR radiation protection program are presented. Programmatic areas and functions important in preventing, responding to, and minimizing radiological incidents and the importance of performing effective incident evaluations and investigations are described. Elements that are integral in ensuring continuous program improvements are emphasized throughout the text.

  16. Operator Support System for Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Operator Support System for Pressurized Water Reactor (OSSPWR) has been developed under the sponsorship of IAEA from August 1994. The project is being carried out by the Department of Engineering Physics, Tsinghua University, Beijing, China. The Design concepts of the operator support functions have been established. The prototype systems of OSSPWR has been developed as well. The primary goal of the project is to create an advanced operator support system by applying new technologies such as artificial intelligence (AI) techniques, advanced communication technologies, etc. Recently, the advanced man-machine interface for nuclear power plant operators has been developed. It is connected to the modern computer systems and utilizes new high performance graphic displays. (author). 6 refs, 4 figs

  17. Light water reactor lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

  18. Light water reactor lower head failure analysis

    International Nuclear Information System (INIS)

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response

  19. [Removal of heavy metals from extract of Angelica sinensis by EDTA-modified chitosan magnetic adsorbent].

    Science.gov (United States)

    Ren, Yong; Sun, Ming-Hui; Peng, Hong; Huang, Kai-Xun

    2013-11-01

    The concentrations of heavy metals in the extracting solutions of traditional Chinese medicine are usually very low. Furthermore, a vast number of organic components contained in the extracting solutions would be able to coordinate with heavy metals, which might lead to great difficulty in high efficient removal of them from the extracting solutions. This paper was focused on the removal of heavy metals of low concentrations from the extracting solution of Angelica sinensis by applying an EDTA-modified chitosan magnetic adsorbent (EDTA-modified chitosan/SiO2/Fe3O4, abbreviated as EDCMS). The results showed that EDCMS exhibited high efficiency for the removal of heavy metals, such as Cu, Cd and Pb, e.g. the removal percentage of Cd and Pb reached 90% and 94.7%, respectively. Besides, some amounts of other heavy metals like Zn and Mn were also removed by EDCMS. In addition, the total solid contents, the amount of ferulic acid and the HPLC fingerprints of the extracting solution were not changed significantly during the heavy metal removal process. These results indicate that EDCMS may act as an applicable and efficient candidate for the removal of heavy metals from the extracting solution of A. sinensis.

  20. Process for the preparation of ammonia and heavy water

    International Nuclear Information System (INIS)

    A process for the production of ammonia and heavy water comprises the steps of enriching a flow of water with deuterium in a monothermal isotropic process; supplying a first portion of the deuterium-enriched water to a heavy water preparation plant to produce heavy water and hydrogen; storing a second portion of the deuterium-enriched water substantially without interruption during the colder half of a year; electrolytically dissociating the stored deuterium-enriched water substantially without interruption during the wamer half of a year to form hydrogen; storing a portion of the electrolytically-produced hydrogen during said warmer half of a year while supplying the remainder to a synthesis circuit of a synthesizing plant and subsequently supplying the stored hydrogen to the synthesis circuit during said colder half of a year; removing some of the synthesis gas mixture from the synthesis circuit of the synthesizing plant; burning the removed synthesis gas mixture with air to produce a mixture consisting mainly of water and nitrogen; thereafter condensing and separating the water from the mixture of water and nitrogen; supplying the nitrogen of the mixture of water and nitrogen, the hydrogen from the heavy water preparation plant and the electrolytically-produced hydrogen to the synthesis circuit of the synthesizing plant to produce ammonia; and collecting deuterium-depleted water resulting from said burning step and feeding the collected deuterium-depleted water into the monothermal process

  1. Magnetic filtration of heavy metals containing waters

    International Nuclear Information System (INIS)

    The high-intensity magnetic separation is applied above all in the beneficiation of fine-grained weakly magnetic ores, but also in the treatment of industrial, especially metallurgical and mining waters as well as of wastewaters from nuclear power stations. Similarly, it can be used in the field of geothermal energy supply and gases filtration. The magnetic separation or filtration respectively, directly enables the treatment of waters contaminated by solid ferromagnetic and paramagnetic particles. The magnetic filtration can remove heavy metals ions and even the oil substances by means of magnetic sorbents or special additives. The filtration of solid magnetic particles can be carried out in matrix-less and matrix separators. On the basis of mathematical description of particles dynamics and hydrodynamic conditions of suspension flow which resulted in the determination of geometrical parameters of separating zone the design of matrix-less magnetic separator was carried out. A strong, high-intensity magnetic field was created by means of a superconductive magnetic circuit. It was found out that for the achievement of optimal technological parameters during the magnetic separation of solid particles with grain size under 40 mm, the maximal solids concentration is to be 200 g/L. The design of matrix parameters and selection of inductive filling resides in theoretical considerations as well as in experimental works. Under laboratory condition the influence of following parameters on magnetic filtration process have been observed: the diameter of inductive ferromagnetic balls, the thickness of filtration layer, the intensity of magnetic field, the flow velocity of suspension, the density of suspension, the grain size of solids and the temperature of suspension. It was found that a spatial arrangement of inductive bodies in filtration layer influences not only the velocity of suspension flow but also a room size for catching of magnetic particles. The acting of magnetic

  2. SPECIATION OF HEAVY METALS AT WATER-SEDIMENT INTERFACE

    Directory of Open Access Journals (Sweden)

    Chiara Ferronato

    2013-09-01

    Full Text Available The objective of the study was to understand the equilibrium relationship between the heavy metals concentrations in superficial water and pore water. At  water-sediment interface, the equilibrium rapidly changed and it is influenced by chemico-physical parameters of aquatic ecosystems. The hydraulic safety of Bologna plain (North Italy depends on network of artificial canals and they are related with natural rivers of Reno basin (Reno river and its tributaries. The natural and artificial water courses flowed in agricultural, urban and industrial land. The heavy metals concentration in water and sediment discriminated the human pressure on the land and their spatial distribution in sediment could predict the hazard of pollution in aquatic ecosystems. We compared the heavy metals concentrations in pore water and superficial water determined in natural rivers and artificial canals, and more pollution in artificial canals than natural rivers was found. Furthermore, the coefficient of partition (log Kd between water and sediments was calculated to evaluate the bioavailability of heavy metals adsorbed on the sediments. The heavy metals extracted in deionised water at equilibrium after 16 h showed higher concentrations than those determined directly on water samples.

  3. Dynamic calculations of pressurized water reactor internals

    International Nuclear Information System (INIS)

    A mathematical model is briefly described for the calculation of oscillations in the WWER-440 reactor internals. The model was developed for improved safety of the type of reactors. It allows calculating vibrations resistance of reactor components, mainly during accidents, such as loss of coolant accidents. Some results are given of the calculation of forces acting in the rupture of the reactor inlet and outlet pipes. (Z.M.)

  4. Radioactive waste management practices with KWU-boiling water reactors

    International Nuclear Information System (INIS)

    A Kraftwerk Union boiling water reactor is used to demonstrate the reactor auxiliary systems which are applied to minimize the radioactive discharge. Based on the most important design criteria the philosophy and function of the various systems for handling the off-gas, ventilation air, waste water and concentrated waste are described. (orig.)

  5. Alkali metal and ammonium chlorides in water and heavy water (binary systems)

    CERN Document Server

    Cohen-Adad, R

    1991-01-01

    This volume surveys the data available in the literature for solid-fluid solubility equilibria plus selected solid-liquid-vapour equilibria, for binary systems containing alkali and ammonium chlorides in water or heavy water. Solubilities covered are lithium chloride, sodium chloride, potassium chloride, rubidium chloride, caesium chloride and ammonium chloride in water and heavy water.

  6. Spatiotemporal Analysis of Heavy Metal Water Pollution in Transitional China

    OpenAIRE

    Huixuan Li; Yingru Li; Ming-Kuo Lee; Zhongwei Liu; Changhong Miao

    2015-01-01

    China’s socioeconomic transitions have dramatically accelerated its economic growth in last three decades, but also companioned with continuous environmental degradation. This study will advance the knowledge of heavy metal water pollution in China from a spatial–temporal perspective. Specifically, this study addressed the following: (1) spatial patterns of heavy metal water pollution levels were analyzed using data of prefecture-level cities from 2004 to 2011; and (2) spatial statistical...

  7. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - I: DUPIC Fuel Fabrication Cost

    International Nuclear Information System (INIS)

    A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC (∼49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC

  8. Study of the prospects for heavy water production via laser isotope separation. Final report, 1 April--30 September 1976

    International Nuclear Information System (INIS)

    The rationale for ''cheap'' heavy water is discussed with emphasis on the economic, safety, and arms control implications of the widespread adoption of pressure-tube, heavy water moderated and cooled CANDU reactors, and variations thereof. Three classes of vibrational-photochemical laser processes are considered in detail, i.e., hydrogen halide-unsaturated hydrocarbon addition reactions, isotopically selective photoadsorption and photodesorption, and selective two-step molecular dissociation. General remarks are made concerning the economic viability of these techniques vis a vis hydrogen sulfide/water chemical exchange

  9. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    Energy Technology Data Exchange (ETDEWEB)

    Widiawati, Nina, E-mail: nina-widiawati28@yahoo.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Jalan Ganesha 10, Bandung (Indonesia)

    2015-09-30

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from −0.6695443 % at BOC to −0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  10. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    Science.gov (United States)

    Widiawati, Nina; Su'ud, Zaki

    2015-09-01

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from -0.6695443 % at BOC to -0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

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

    International Nuclear Information System (INIS)

    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

  12. Introduction to reactor internal materials for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

  13. Introduction to reactor internal materials for pressurized water reactor

    International Nuclear Information System (INIS)

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,

  14. Decommissioning a nuclear reactor. [Water Boiler Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    Montoya, G.M.

    1991-01-01

    The process of decommissioning a facility such as a nuclear reactor or reprocessing plant presents many waste management options and concerns. Waste minimization is a primary consideration, along with protecting a personnel and the environment. Waste management is complicated in that both radioactive and chemical hazardous wastes must be dealt with. This paper presents the general decommissioning approach of a recent project at Los Alamos. Included are the following technical objectives: site characterization work that provided a thorough physical, chemical, and radiological assessment of the contamination at the site; demonstration of the safe and cost-effective dismantlement of a highly contaminated and activated nuclear-fuelded reactor; and techniques used in minimizing radioactive and hazardous waste. 12 figs.

  15. Dithiocarbamate-modified starch derivatives with high heavy metal adsorption performance.

    Science.gov (United States)

    Xiang, Bo; Fan, Wen; Yi, Xiaowei; Wang, Zuohua; Gao, Feng; Li, Yijiu; Gu, Hongbo

    2016-01-20

    In this work, three types of dithiocarbamate (DTC)-modified starch derivatives including DTC starch (DTCS), DTC enzymolysis starch (DTCES) and DTC mesoporous starch (DTCMS) were developed, which showed the significant heavy metal adsorption performance. The adsorption ability of these three DTC modified starch derivatives followed the sequences: DTCMS>DTCES>DTCS. In single metal aqueous solutions, the uptake amount of heavy metal ions onto the modified starches obeyed the orders: Cu(II)>Ni(II)>Cr(VI)>Zn(II)>Pb(II). The adsorption mechanism was proved by the chelating between DTC groups and heavy metal ions through the pH effect measurements. A monolayer adsorption of Langmuir isotherm model for the adsorption of Cu(II) onto DTCMS was well fitted rather than the multilayer adsorption of Freundlich isotherm model. The adsorption kinetics of Cu(II) onto starch derivatives was found to be fit well with the pseudo-second-order model. Additionally, in the presence of EDTA, the adsorption ability and uptake amount of heavy metal ions onto these three DTC modified starch derivatives is identical with the results obtained in the absence of EDTA.

  16. Modification of the Argonne tandem

    International Nuclear Information System (INIS)

    For nuclear structure experiments with heavy ions it is necessary to have ion energies in excess of 5 MeV per nucleon. At the Argonne tandem FN accelerator this was accomplished by the addition of a superconducting linac. Modifications of the FN tandem to improve the performance of the pair is described

  17. HIBALL - a conceptual heavy ion beam driven fusion reactor study. Vol. 2

    International Nuclear Information System (INIS)

    A preliminary concept for a heavy-ion beam driven inertial confinement fusion power plant is presented. The high repetition rate of the RF accelerator driver is utilized to serve four reactor chambers alternatingly. In the chambers a novel first-wall protection scheme is used. At a target gain of 83 the total net electrical output is 3.8 GW. The recirculating power fraction is below 15%. The main goal of the comprehensive HIBALL study (which is continuing) is to demonstrate the compatibility of the design of the driver, the target and the reactor chambers. Though preliminary, the present design is essentially self-consistent. Tentative cost estimates are given. The costs compare well with those found in similar studies on other types of fusion reactors. (orig.)

  18. HIBALL - a conceptual heavy ion beam driven fusion reactor study. Vol. 1

    International Nuclear Information System (INIS)

    A preliminary concept for a heavy-ion beam driven inertial confinement fusion power plant is presented. The high repetition rate of the RF accelerator driver is utilized to serve four reactor chambers alternatingly. In the chambers a novel first-wall protection scheme is used. At a target gain of 83 the total net electrical output is 3.8 GW. The recirculating power fraction is below 15%. The main goal of the comprehensive HIBALL study (which is continuing) is to demonstrate the compatibility of the design of the driver, the target and the reactor chambers. Though preliminary, the present dessign is essentially self-consistent. Tentative cost estimates are given. The costs compare well with those found in similar studies on other types of fusion reactors. (orig.)

  19. Nonthermal plasma reactors for the production of light hydrocarbon olefins from heavy oil

    Directory of Open Access Journals (Sweden)

    G. Prieto

    2003-03-01

    Full Text Available During the last decade, nonthermal plasma technology was applied in many different fields, focusing attention on the destruction of harmful compounds in the air. This paper deals with nonthermal plasma reactors for the conversion of heavy oil into light hydrocarbon olefins, to be employed as gasoline components or to be added in small amounts for the catalytic reduction of nitrogen oxide compounds in the treatment of exhaust gas at power plants. For the process, the plate-plate nonthermal plasma reactor driven by AC high voltage was selected. The reactor was modeled as a function of parameter characteristics, using the methodology provided by the statistical experimental design. The parameters studied were gap distance between electrodes, carrier gas flow and applied power. Results indicate that the reactions occurring in the process of heavy oil conversion have an important selective behavior. The products obtained were C1-C4 hydrocarbons with ethylene as the main compound. Operating the parameters of the reactor within the established operative window of the system and close to the optimum conditions, efficiencies as high as 70 (mul/joule were obtained. These values validate the process as an in-situ method to produce light olefins for the treatment of nitrogen oxides in the exhaust gas from diesel engines.

  20. Reliability assurance programme guidebook for advanced light water reactors

    International Nuclear Information System (INIS)

    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

  1. Water desalination by a fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    The great need for potable water in the world motivated the International Atomic Energy Agency (IAEA) to study the feasibility of nuclear seawater desalination. The consensus reached is that nuclear desalination is technically feasible, though cost and social acceptability are recognized as major problems to overcome. Here an inherently safe reactor with reduced cost is proposed to overcome these barriers. The reactor is a simple small modular nuclear reactor based on fluidized bed concept with passive cooling characteristics. (orig.)

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

    International Nuclear Information System (INIS)

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

  3. Light water reactor piping system damping

    International Nuclear Information System (INIS)

    In this paper, based on a detailed evaluation and screening of existing damping data, a set of damping values are recommended for light water reactor piping systems. A multivariate regression model was used to identify the significant physical and response characteristics of piping systems. Although initially several experimental biases were identified that help explain the large variability in the existing data, these were ignored and only physical attributes were considered for the final recommendations. Of these twenty-two initial variables, only six were identified as being important to energy dissipation. Since the existing data is incomplete for certain variables, the identified parameters are not an exhaustive set. A regression analysis can only identify those parameters as significant that have a sufficient number and a wide spectrum of data points. Making several conservation assumptions, the six variable damping prediction equation was reduced to a damping table with two parameters: Response Level and Diameter. Pipe diameter is a convenient simple characteristic to represent system stiffness and hence support/pipe interaction, which tends to be a significant source of energy dissipation in piping systems

  4. Fundamentals of boiling water reactor systems

    International Nuclear Information System (INIS)

    The reactor assembly consists of the reactor vessel, its internal components of the core, shroud, steam separator, dryer assemblies, feedwater spargers, internal recirculation pumps and control rod drive housings. Connected to the steam lines are the pressure relief valves which protect the pressure boundary from damage due to overpressure. (orig./TK)

  5. Anion-intercalated layered double hydroxides modified test strips for detection of heavy metal ions.

    Science.gov (United States)

    Wang, Nan; Sun, Jianchao; Fan, Hai; Ai, Shiyun

    2016-02-01

    In this work, a novel approach for facile and rapid detection of heavy metal ions using anion-intercalated layered double hydroxides (LDHs) modified test strips is demonstrated. By intercalating Fe(CN)6(4-) or S(2-) anions into the interlayers of LDHs on the filter paper, various heavy metal ions can be easily detected based on the color change before and after reaction between the anions and the heavy metal ions. Upon the dropping of heavy metal ions solutions to the test strips, the colors of the test strips changed instantly, which can be easily observed by naked eyes. With the decrease of the concentration, the color depth changed obviously. The lowest detection concentration can be up to 1×10(-6) mol L(-1). Due to the easily intercalation of anions into the interlayer of the LDHs on test trips, this procedure provides a general method for the construction of LDHs modified test strips for detection of heavy metal ions. The stability of the prepared test strips is investigated. Furthermore, all the results were highly reproducible. The test strips may have potential applications in environmental monitoring fields. PMID:26653453

  6. Overview of environmental materials degradation in light-water reactors

    International Nuclear Information System (INIS)

    This report provides a brief overview of analyses and conclusions reported in published literature regarding environmentally induced degradation of materials in operating light-water reactors. It is intended to provide a synopsis of subjects of concern rather than to address a licensing basis for any newly discovered problems related to reactor materials

  7. Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

    International Nuclear Information System (INIS)

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive ''box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs

  8. Modified Woods-Saxon Potential for Heavy-Ion Fusion Reaction

    Institute of Scientific and Technical Information of China (English)

    TIAN Jun-Long; WANG Ning; LI Zhu-Xia

    2007-01-01

    A modified Woods-Saxon(MWS)potential is proposed for describing nucleus-nucleus interaction based on the Skyrme energy-density functional approach.Fusion barriers for a large number of fusion reactions from light to heavy systems can be described well with this potential.The suitable incident energies for fusion reactions leading to superheavy nuclei are also explored.It seems to US that the MWS potential is useful for selecting the suitable incident energies for fusion reactions for producing super-heavy nuclei.

  9. Modified phase rule for inelastic scattering of heavy ions near the Coulomb barrier

    Energy Technology Data Exchange (ETDEWEB)

    Wohlfarth, D.; Hentschel, E.; Hersch, G.U.; Grambole, D.; Man' ko, V.I.; Sakuta, S.B.; Thomas, H.J.; Chuev, V.I.

    1983-12-01

    We propose a semiclassical model of heavy-ion inelastic scattering in which the difference in the entry and exit trajectories is taken into account. This model permits explanation of most of the experimentally observed deviations from the 180/sup 0/ phase rule formulated for elastic and inelastic scattering of heavy ions at energy near the Coulomb barrier. It is shown that in those cases when even the modified phase rule which follows from the semiclassical model does not agree with experiment, a two-stage mechanism gives a substantial contribution to inelastic scattering.

  10. Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Kassner, T. F.; Ruther, W. E.; Shack, W. J.; Smith, J. L.; Soppet, W. K.; Strain; R. V. (Energy Technology); ( APS-USR)

    1999-10-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1998 to December 1998. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. Fatigue tests have been conducted to determine the crack initiation and crack growth characteristics of austenitic SSs in LWR environments. Procedures are presented for incorporating the effects of reactor coolant environments on the fatigue life of pressure vessel and piping steels. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in helium at 289 C in the Halden reactor. The results have been used to determine the influence of alloying and impurity elements on the susceptibility of these steels to irradiation-assisted stress corrosion cracking. Fracture toughness J-R curve tests were also conducted on two heats of Type 304 SS that were irradiated to {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. Crack-growth-rate tests have been conducted on compact-tension specimens of Alloys 600 and 690 under constant load to evaluate the resistance of these alloys to stress corrosion cracking in LWR environments.

  11. Shielding designs for pressurized water reactors in France

    Energy Technology Data Exchange (ETDEWEB)

    Champion, G.; Forestier, J.; Vergnaud, T.

    1986-07-01

    The efforts made by Electricite de France to reduce exposure from the two-component neutron-gamma radiation fields inside the pressurized water reactor (PWR) building are described. Most of the attention had been focused on the problem of neutron exposure relative to the problem of achieving a highly efficient confinement within the reactor cavity and the state of the art of personnel neutron dosimetry. A description of the general neutron calculation scheme that links the characteristics of the neutron fields escaping from the reactor vessel to the dose equivalent rate cartographies inside the reactor building is provided.

  12. Spatiotemporal Analysis of Heavy Metal Water Pollution in Transitional China

    Directory of Open Access Journals (Sweden)

    Huixuan Li

    2015-07-01

    Full Text Available China’s socioeconomic transitions have dramatically accelerated its economic growth in last three decades, but also companioned with continuous environmental degradation. This study will advance the knowledge of heavy metal water pollution in China from a spatial–temporal perspective. Specifically, this study addressed the following: (1 spatial patterns of heavy metal water pollution levels were analyzed using data of prefecture-level cities from 2004 to 2011; and (2 spatial statistical methods were used to examine the underlying socioeconomic and physical factors behind water pollution including socioeconomic transitions (industrialization, urbanization, globalization and economic development, and environmental characteristic (natural resources, hydrology and vegetation coverage. The results show that only Cr pollution levels increased over the years. The individual pollution levels of the other four heavy metals, As, Cd, Hg, and Pb, declined. High heavy metal water pollution levels are closely associated with both anthropogenic activities and physical environments, in particular abundant mineral resources and industrialization prosperity. On the other hand, economic development and urbanization play important roles in controlling water pollution problems. The analytical findings will provide valuable information for policy-makers to initiate and adjust protocols and strategies for protecting water sources and controlling water pollution; thus improving the quality of living environments.

  13. Dependence of heavy metal burnup on nuclear data libraries for fast reactors

    CERN Document Server

    Ohki, S

    2003-01-01

    Japan Nuclear Cycle Development Institute (JNC) is considering the highly burnt fuel as well as the recycling of minor actinide (MA) in the development of commercialized fast reactor cycle systems. Higher accuracy in burnup calculation is going to be required for higher mass plutonium isotopes ( sup 2 sup 4 sup 0 Pu, etc.) and MA nuclides. In the framework of research and development aiming at the validation and necessary improvements of fast reactor burnup calculation, we investigated the differences among the burnup calculation results with the major nuclear data libraries: JEF-2.2, ENDF/B-VI Release 5, JENDL-3.2, and JENDL-3.3. We focused on the heavy metal nuclides such as plutonium and MA in the central core region of a conventional sodium-cooled fast reactor. For main heavy metal nuclides ( sup 2 sup 3 sup 5 U, sup 2 sup 3 sup 8 U, sup 2 sup 3 sup 9 Pu, sup 2 sup 4 sup 0 Pu, and sup 2 sup 4 sup 1 Pu), number densities after 1-cycle burnup did not change over one or two percent. Library dependence was re...

  14. Capital Cost: Pressurized Water Reactor Plant Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139-MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume includes in addition to the foreword and summary, the plant description and the detailed cost estimate.

  15. Hydrogen production by water dissociation from a nuclear reactor

    International Nuclear Information System (INIS)

    This memento presents the production of hydrogen by water decomposition, the energy needed for the electrolysis, the thermochemical cycles for a decomposition at low temperature and the possible nuclear reactors associated. (A.L.B.)

  16. RB Research nuclear reactor, Annual report for 1994, I - III

    International Nuclear Information System (INIS)

    Report on RB reactor operation during 1994 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization as well as operation of the VAX-8250 computer

  17. Modified Smith Predictor Based Control Of Cascaded Chemical Reactor

    Directory of Open Access Journals (Sweden)

    Binu P. Mathew

    2014-04-01

    Full Text Available A cascade control with modified smith predictor is used for controlling an open loop unstable time delay process. It has three controllers, one is for servo response other two are for regulatory response. For two disturbance rejection controllers an analytical design method is used by proposing closed loop complementary sensitivity function. These two controllers are PID controller cascaded with second order lead/lag filter. Setpoint tracking controller is designed by using direct synthesis method. The main advantage of this control scheme is that the servo response can be decoupled from the regulatory response.

  18. SPECIATION OF HEAVY METALS AT WATER-SEDIMENT INTERFACE

    OpenAIRE

    Chiara Ferronato; Livia Vittori Antisari; Monica Marianna Modesto; Gilmo Vianello

    2013-01-01

    The objective of the study was to understand the equilibrium relationship between the heavy metals concentrations in superficial water and pore water. At  water-sediment interface, the equilibrium rapidly changed and it is influenced by chemico-physical parameters of aquatic ecosystems. The hydraulic safety of Bologna plain (North Italy) depends on network of artificial canals and they are related with natural rivers of Reno basin (Reno river and its tributaries). The natural and artificial w...

  19. Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (CE-STS) is a generic document prepared by the US NRC for use in the licensing process of current Combustion Engineering Pressurized Water Reactors. The CE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  20. Standard Technical Specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements

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

    International Nuclear Information System (INIS)

    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)

  2. Energy conservation and management strategies in Heavy Water Plants

    International Nuclear Information System (INIS)

    In the competitive industrial environment it is essential that cost of the product is kept at the minimum possible. Energy conservation is an important aspect in achieving this as energy is one of the key recourses for growth and survival of industry. The process of heavy water production being very complex and energy intensive, Heavy Water board has given a focussed attention for initiating various measures for reducing the specific energy consumption in all the plants. The initiative resulted in substantial reduction in specific energy consumption and brought in savings in cost. The cumulative reduction of specific energy consumption has been over 30% over the last seven years and the total savings for the last three years on account of the same has been about Rs. 190 crore. The paper describes the strategies adopted in the heavy water plants for effecting the above achievements. The paper covers the details of some of the energy saving schemes carried out at different heavy water plants through case studies. The case studies of schemes implemented at HWPs are general in nature and is applicable for any other industry. The case studies cover the modifications with re-optimisation of the process parameters, improvements effected in utility units like refrigeration and cooling water systems, improvements in captive power plant cycle and improved recycle scheme for water leading to reduced consumptions. The paper also mentions the innovative ammonia absorption refrigeration with improved coefficient of performance and HWB's efforts in development of the system as an integrated unit of the ammonia water deuterium exchange process for heavy water production. HWB also has taken up R and D on various other schemes for improvements in energy consumption for future activities covering utilisation of low grade energy for generation of refrigeration. (author)

  3. Using biopolymers to remove heavy metals from soil and water

    Energy Technology Data Exchange (ETDEWEB)

    Krishnamurthy, S.; Frederick, R.M.

    1993-11-19

    Chemical remediation of soil may involve the use of harsh chemicals that generate waste streams, which may adversely affect the soil's integrity and ability to support vegetation. This article reviews the potential use of benign reagents, such as biopolymers, to extract heavy metals. The biopolymers discussed are chitin and chitosan, modified starch, cellulose, and polymer-containing algae. (Copyright (c) Remediation 1994.)

  4. Transference of know-how for the fabrication of heavy components for nuclear power reactors

    International Nuclear Information System (INIS)

    1) Heavy components for nuclear power reactors. Reactor pressure vessels with total weight of 540 tons; steam generators: heat exchangers with U-type tube bundles, total weight 420 tons. 2) Choice of know-how recipient. Technical criteria, i.e. manufacturing facilities, existing quality assurance system, location of the workshops, possibilities for training, infrastructures. 3. Measures for transferring know-how to a newly established company. Planning and erection of the factory: organisational set up of the company; personnel selection and training; transfer of documentation; transfer of know-how that cannot be transferred in a written form. 4) Contracts for assuring the transfer of know-how. Stipulation of mutual rights and obligations of the know-how owner and receiver in individual contracts: engineering services contract, technical information contract, personnel training contract, license contract. (orig.)

  5. Nuclear analysis of the heavy-ion-beam-driven fusion reactor HIBALL

    International Nuclear Information System (INIS)

    A detailed three-dimensional Monte Carlo nuclear analysis is presented for the heavy-ion-beam-driven reactor HIBALL. Neutron target interactions leading to neutron multiplication, spectrum softening, and gamma production are included in the model. A 0.66m-thick blanket cooleb by Pb83 Li 17 reduces the radiation damage in the HT-9 ferritic steel first wall to 2.7 dpa/full power year, allowing it to last the whole life of the plant. The overall tritium breeding ratio and the overall energy multiplication are 1.25 and 1.27, respectively. The four reactor cavities in the HIBALL power plant yield a total thermal power of 10 200 MW(thermal)

  6. Improving economics and safety of water cooled reactors. Proven means and new approaches

    International Nuclear Information System (INIS)

    Nuclear power plants (NPPs) with water cooled reactors [either light water reactors (LWRs) or heavy water reactors (HWRs)] constitute the large majority of the currently operating plants. Water cooled reactors can make a significant contribution to meeting future energy needs, to reducing greenhouse gas emissions, and to energy security if they can compete economically with fossil alternatives, while continuing to achieve a very high level of safety. It is generally agreed that the largest commercial barrier to the addition of new nuclear power capacity is the high capital cost of nuclear plants relative to other electricity generating alternatives. If nuclear plants are to form part of the future generating mix in competitive electricity markets, capital cost reduction through simplified designs must be an important focus. Reductions in operating, maintenance and fuel costs should also be pursued. The Department of Nuclear Energy of the IAEA is examining the competitiveness of nuclear power and the means for improving its economics. The objective of this TECDOC is to emphasize the need, and to identify approaches, for new nuclear plants with water cooled reactors to achieve competitiveness while maintaining high levels of safety. The cost reduction methods discussed herein can be implemented into plant designs that are currently under development as well as into designs that may be developed in the longer term. Many of the approaches discussed also generally apply to other reactor types (e.g. gas cooled and liquid metal cooled reactors). To achieve the largest possible cost reductions, proven means for reducing costs must be fully implemented, and new approaches described in this document should be developed and implemented. These new approaches include development of advanced technologies, increased use of risk-informed methods for evaluating the safety benefit of design features, and international consensus regarding commonly acceptable safety requirements that

  7. HIBALL-II - an improved conceptual heavy ion beam driven fusion reactor study

    International Nuclear Information System (INIS)

    An improved design of the HIBALL inertial-confinement fusion power station is presented. The new RF-linac based heavy ion driver has improved concepts for beam stacking, bunching and final focusing. The new target design takes into account radiation transport effects in a coarse approximation. The system of four reactors with a net total output of 3.8 GW electric is essentially the same as described earlier, however, progress in the analysis has enhanced its credibility and self-consistency. Considerations of environmental and safety aspects and cost estimates are given. (orig.)

  8. Peptidolytic microbial community of methanogenic reactors from two modified UASBs of brewery industries

    Directory of Open Access Journals (Sweden)

    C. Díaz

    2010-10-01

    Full Text Available We studied the peptide-degrading anaerobic communities of methanogenic reactors from two mesophilic full-scale modified upflow anaerobic sludge blanket (UASB reactors treating brewery wastewater in Colombia. Most probable number (MPN counts varied between 7.1 x 10(8 and 6.6 x 10(9 bacteria/g volatile suspended solids VSS (Methanogenic Reactor 1 and 7.2 x 10(6 and 6.4 x 10(7 bacteria/g (VSS (Methanogenic Reactor 2. Metabolites detected in the highest positive MPN dilutions in both reactors were mostly acetate, propionate, isovalerate and, in some cases, negligible concentrations of butyrate. Using the highest positive dilutions of MPN counts, 50 dominant strains were isolated from both reactors, and 12 strains were selected for sequencing their 16S rRNA gene based on their phenotypic characteristics. The small-subunit rRNA gene sequences indicated that these strains were affiliated to the families Propionibacteriaceae, Clostridiaceae and Syntrophomonadaceae in the low G + C gram-positive group and Desulfovibrio spp. in the class d-Proteobacteria. The main metabolites detected in the highest positive dilutions of MPN and the presence of Syntrophomonadaceae indicate the effect of the syntrophic associations on the bioconversion of these substrates in methanogenic reactors. Additionally, the potential utilization of external electron acceptors for the complete degradation of amino acids by Clostridium strains confirms the relevance of these acceptors in the transformation of peptides and amino acids in these systems.

  9. Utilizing heavy metal-laden water hyacinth biomass in vermicomposting.

    Science.gov (United States)

    Tereshchenko, Natalya N; Akimova, Elena E; Pisarchuk, Anna D; Yunusova, Tatyana V; Minaeva, Oksana M

    2015-05-01

    We studied the efficiency of water treatment by water hyacinth (Eichhornia crassipes) from heavy metals (Zn, Cd, Pb, Cu), as well as a possibility of using water hyacinth biomass obtained during treatment for vermicomposting by Eisenia fetida and the vermicompost quality in a model experiment. The results showed that the concentration of heavy metals in the trials with water hyacinth decreased within 35 days. We introduced water hyacinth biomass to the organic substrate for vermicomposting, which promoted a significant weight gain of earthworms and growth in their number, as well as a 1.5- to 3-fold increase in coprolite production. In the trial with 40 % of Eichhornia biomass in the mixture, we observed a 26-fold increase in the number and a 16-fold weight gain of big mature individuals with clitellum; an increase in the number of small individuals 40 times and in the number of cocoons 140 times, as compared to the initial substrate. The utilization of water hyacinth biomass containing heavy metals in the mixture led to a 10-fold increase in the number of adult individuals and cocoons, which was higher than in control. We found out that adding 10 % of Eichhornia biomass to the initial mixture affected slightly the number of microorganisms and their species diversity in the vermicompost. Adding Eichhornia biomass with heavy metals reduced the total number of microorganisms and sharply diminished their species diversity. In all trials, adding water hyacinth in the mixture for vermicomposting had a positive impact on wheat biometric parameters in a 14-day laboratory experiment, even in the trial with heavy metals. PMID:25501861

  10. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    Science.gov (United States)

    Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-01

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  11. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    International Nuclear Information System (INIS)

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated

  12. Light Water Reactor Sustainability Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80% below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you

  13. Water chemistry control to meet the advanced design and operation of light water reactors

    International Nuclear Information System (INIS)

    Water chemistry control is one of the key technologies to establish safe and reliable operation of nuclear power plants. The road maps on R and D plans for water chemistry of nuclear power systems in Japan have been proposed along with promotion of R and D related water chemistry improvement for the advanced application of light water reactors (LWRs). The technical trends were divided into four categories, dose rate reduction, structural integrity, fuel integrity and radioactive waste reduction, and latest technical break through for each category was shown for the advanced application of LWRs. At the same time, the technical break through and the latest movements for regulation of water chemistry were introduced for each of major organizations related to nuclear engineering in the world. The conclusions were summarized as follows; 1. Water chemistry improvements might contribute to achieve the advanced application of LWRs, while water chemistry should be often changed to achieve the advanced application of LWRs. 2. Only one solution for water chemistry control was not obtained for achieving the advanced application of LWRs, but miscellaneous solutions were possible for achieving one. Optimal water chemistry control was desired for having the good practices for satisfying multi-targets at the same time and it was much affected by the plant unique systems and operational history. 3. That meant it was difficult to determine water chemistry regulation targets for achieving application of LWRs but it was necessary to prepare suitable guideline for good achievement of application of LWRs. That meant the guideline should be recommendation for good practice in the plant. 4. The water chemistry guide line should be modified along with progress of plant operation and water chemistry and related technologies. (author)

  14. High fuel burn-up and nonproliferation in PWR-type reactor on the basis of modified Th-fuel

    International Nuclear Information System (INIS)

    Neutronics-physical characteristics of the fuel lattice of a PWR-type reactor cooled by light water and by a mixture of light and heavy water have been analyzed. Th-fuel containing an essential amount of 231Pa and 232U is used, which allows an increase in fuel burn-up by a factor of 2-5 compared with that of traditional oxide uranium fuel with light water. It is important to underline that this is attained under the negative coolant density reactivity effect using cross sections of 231Pa and 232U from the updated JENDL-3.2 nuclear library. This radical increase of fuel burn-up is accompanied by a small change of reactivity during fuel irradiation (K∞=1.1 / 1.0), that favorably affects safety parameters of the reactor operation. A considerable percentage of 232U in fuel, and consequently in U, is a strong barrier against the proliferation of such weapon nuclide as 233U. (authors)

  15. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  16. Membrane reactor for water detritiation: a parametric study on operating parameters

    Energy Technology Data Exchange (ETDEWEB)

    Mascarade, J.; Liger, K.; Troulay, M.; Perrais, C. [CEA, DEN, DTN/STPA/LIPC, Centre de Cadarache, Saint-Paul-lez-Durance (France); Joulia, X.; Meyer, X.M. [Universite de Toulouse, INPT, UPS, Laboratoire de Genie Chimique, Toulouse (France); CNRS, Laboratoire de Genie Chimique, Toulouse (France)

    2015-03-15

    This paper presents the results of a parametric study done on a single stage finger-type packed-bed membrane reactor (PBMR) used for heavy water vapor de-deuteration. Parametric studies have been done on 3 operating parameters which are: the membrane temperature, the total feed flow rate and the feed composition through D{sub 2}O content variations. Thanks to mass spectrometer analysis of streams leaving the PBMR, speciation of deuterated species was achieved. Measurement of the amounts of each molecular component allowed the calculation of reaction quotient at the packed-bed outlet. While temperature variation mainly influences permeation efficiency, feed flow rate perturbation reveals dependence of conversion and permeation properties to contact time between catalyst and reacting mixture. The study shows that isotopic exchange reactions occurring on the catalyst particles surface are not thermodynamically balanced. Moreover, the variation of the heavy water content in the feed exhibits competition between permeation and conversion kinetics.

  17. A water cooled, lithium lead breeding blanket for a DEMO fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Rieger, M.; Biggio, M.; Farfaletti-Casali, F.; Tominetti, S.; Wu, J.; Zucchetti, M. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Labbe, P.; Baraer, L.; Gervaise, G.; Giancarli, L.; Roze, M.; Severi, Y.; Quintric-Bossy, J. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))

    1991-04-01

    The main features of a tritium breeding blanket for a Demonstration Power Reactor involving the eutectic Pb-17Li as liquid breeder and water as coolant are presented. The configuration of the blanket segments and breeder modules as well as their arrangement inside the reactor vacuum vessel are outlined. The main design aspects and the corresponding design limits are reviewed, namely those related to thermomechanics, neutronics, magneto-hydrodynamics, tritium permeation and recovery. First results of safety analysis, in particular those connected with the rupture of a coolant tube in the breeder module are presented and discussed. As a conclusion, the feasibility of the concept look attractive. A problem which requires further investigation is that of the tritium self-sufficiency. It is shown that a net tritium production near to one can be obtained if berylium tiles are placed in front of the plasma, provided that they are cooled by heavy water. (orig.).

  18. Assessment of Heavy Metals Pollution in Yu - Water Copper by Modified BCR Extraction Procedure%BCR法研究铜矿区周边农田重金属形态分布

    Institute of Scientific and Technical Information of China (English)

    汤嘉南; 彭娜; 涂常青; 王似玉; 顾忠林

    2011-01-01

    The three- step BCR(eommunity bureau of reference)sequential extraction procedure method is used to analyse the spociation and bioavailability of heavy metals( Cu, Pb ,Zn, Mn)in agricultural soils around copper mine tailing in Meizhou. Results show that downstream soil around mine is polluted heavily, which is polluted more seri- ously than upstream soil. Soil aronud mine is polluted by Pb mainly. In upstream and downstream soil around mine, Cu, Zn, Mn is dominantly released in the residual fraction. Pb is mainly released in the reducible fraction. It means that in upstream and downstream soil, Pb and Zn has the highest and lowest bioavailability respetivly.%采用BCR(communitybureauofreference)连续提取法对梅县玉水铜矿区周边农田土壤重金属Cu、Pb、zIl和Mn的形态分布及其生物有效性进行了分析.结果表明,矿区下游地区农田土壤重金属污染比上游严重,属于重污染区,而上游土壤属于轻度污染.矿区上游和下游农田土壤中,Pb的污染贡献最大.上游和下游土壤中cu、Zn、Mn都主要分布在残渣态中,Pb主要分布在可还原态.上游和下游土壤中都以Pb的有效性最高,zn的有效性最低.

  19. Selected bibliography on deuterium isotope effects and heavy water

    International Nuclear Information System (INIS)

    In recent years, there has been a great deal of interest in using deuterium and heavy water not only in nuclear industry but also in various fields of basic as well as applied research in physics, chemistry and biology. As a result, the literature is being enriched with a large number of research papers and technical reports published each year. Thus, to enable the scientists to have an easy reference to these works, an endeavour has been made in this selected bibliography, to enlist the publications related to these fields. Since the interest is concerned mainly with heavy water production processes, deuterium isotope effects etc., several aspects (e.g. nuclear) of deuterium have not been covered here. The material in this bibliography which cites 2388 references has been classified under six broad headings, viz. (1) Production of heavy water, (2) Study of deuterium isotope effects, (3) Analysis and Properties of heavy water, (4) Laser Separation of deuterium, (5) Isotopic exchange reactions, and (6) Miscellaneous. The sources of information used for this compilation are chemical abstracts, nuclear science abstracts, INIS Atomindex and also some scattered search through journals and reports available in the B.A.R.C. library. However, in spite of sincere attempts for a wide coverage, no claim is being made towards the exhaustiveness of this bibliography. (author)

  20. Valuation of Embalse Nuclear Power Plant and of heavy water

    International Nuclear Information System (INIS)

    The author describes the Nuclear Power Plant characteristics, the building work, the heavy water valuation criteria and the reasons why he considers that any capital good can be valuated by means of cash-flow. The value of replacement of Embalse Nuclear Power Plant is of U$S 1.593.538.000 (authors)

  1. Embalse nuclear power plant and heavy water valuation

    International Nuclear Information System (INIS)

    The author describes the nuclear power plant characteristics, the building work, the heavy water valuation criteria and the reasons why he considers that any capital good can be valued by the cash-flow method. The Embalse nuclear power plant replacement value is of U$S 1.593.538.000. (author)

  2. Long term assurance of supply of heavy water

    International Nuclear Information System (INIS)

    The answer of Switzerland and Great Britain to a number of questions concerning the long-term assurance of the supply of heavy water are presented. The original problems are seen in the wider context of raw materials supply and its assurance in general. Non-proliferation aspects are touched

  3. Modified kinetic-hydraulic UASB reactor model for treatment of wastewater containing biodegradable organic substrates.

    Science.gov (United States)

    El-Seddik, Mostafa M; Galal, Mona M; Radwan, A G; Abdel-Halim, Hisham S

    2016-01-01

    This paper addresses a modified kinetic-hydraulic model for up-flow anaerobic sludge blanket (UASB) reactor aimed to treat wastewater of biodegradable organic substrates as acetic acid based on Van der Meer model incorporated with biological granules inclusion. This dynamic model illustrates the biomass kinetic reaction rate for both direct and indirect growth of microorganisms coupled with the amount of biogas produced by methanogenic bacteria in bed and blanket zones of reactor. Moreover, the pH value required for substrate degradation at the peak specific growth rate of bacteria is discussed for Andrews' kinetics. The sensitivity analyses of biomass concentration with respect to fraction of volume of reactor occupied by granules and up-flow velocity are also demonstrated. Furthermore, the modified mass balance equations of reactor are applied during steady state using Newton Raphson technique to obtain a suitable degree of freedom for the modified model matching with the measured results of UASB Sanhour wastewater treatment plant in Fayoum, Egypt. PMID:27054727

  4. Modified kinetic-hydraulic UASB reactor model for treatment of wastewater containing biodegradable organic substrates.

    Science.gov (United States)

    El-Seddik, Mostafa M; Galal, Mona M; Radwan, A G; Abdel-Halim, Hisham S

    2016-01-01

    This paper addresses a modified kinetic-hydraulic model for up-flow anaerobic sludge blanket (UASB) reactor aimed to treat wastewater of biodegradable organic substrates as acetic acid based on Van der Meer model incorporated with biological granules inclusion. This dynamic model illustrates the biomass kinetic reaction rate for both direct and indirect growth of microorganisms coupled with the amount of biogas produced by methanogenic bacteria in bed and blanket zones of reactor. Moreover, the pH value required for substrate degradation at the peak specific growth rate of bacteria is discussed for Andrews' kinetics. The sensitivity analyses of biomass concentration with respect to fraction of volume of reactor occupied by granules and up-flow velocity are also demonstrated. Furthermore, the modified mass balance equations of reactor are applied during steady state using Newton Raphson technique to obtain a suitable degree of freedom for the modified model matching with the measured results of UASB Sanhour wastewater treatment plant in Fayoum, Egypt.

  5. Roadmap for implementation of light water reactor decommissioning

    International Nuclear Information System (INIS)

    While decommissioning of Tokai-mura reactor and JATR reactor has already started in Japan, Tsuruga reactor is announced shutdown in 2010 as the first decommissioning of commercial light water reactor (LWR). In 2030s or may be more earlier due to economic reasons, decommissioning of LWRs will take place in succession. Since rational decommissioning needs operating data of individual plants, ample time should be allowed for planning the reactor decommissioning. Committee of Nuclear Power Engineering Cooperation (NUPEC) had identified relevant issues to implement LWR decommissioning and established roadmaps showing fundamental approaches to solve seventeen items categorized in seven areas as action items. Harmonization of policy, regulations and technology development as a whole and reflection of accumulated lessons learned from overseas decommissioning experiences needed further study. (T. Tanaka)

  6. Fatigue and environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Fatigue and environmentally assisted cracking of piping, pressure vessels, and core components in light water reactors (LWRs) are important concerns as extended reactor lifetimes are envisaged. The degradation processes include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or SCC cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Similar cracking has also occurred in upper shell-to-transition cone girth welds in pressurized water reactor (PWR) steam generator vessels. Another concern is failure of reactor-core internal components after accumulation of relatively high fluence, which has occurred in both BWRs and PWRs. Research during the past year focused on (1) fatigue and SCC of ferritic steels used in piping and in steam generator and reactor pressure vessels, (2) role of chromate and sulfate in simulated BWR water in SCC of sensitized Type 304 SS, and (3) irradiation-assisted SCC in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs. Failure after accumulation of relatively high fluence has been attributed to radiation-induced segregation (RIS) of elements such as Si, P, Ni, and Cr. This document provides a summary of research progress in these areas

  7. Replacement steam generators for pressurized water reactors

    International Nuclear Information System (INIS)

    Babcock and Wilcox Canada has developed an Advanced Series steam generator for PWR Systems. This design incorporates all of the features that have contributed to the successful CANDU steam generator performance. This paper presents an overview of the design features and how the overall design relates to the requirements of a PWR reactor system

  8. Modifying the growth morphology of aluminum crystals by magnetic mirror in a thermal plasma reactor

    International Nuclear Information System (INIS)

    Effect of magnetic fields on growth morphology of aluminum crystals was studied in a fluidized bed thermal plasma reactor assisted by magnetic mirrors. Aluminum crystals were precipitated in the reactor using aluminum powder or aluminum-graphite mixture as precursors. The absent of magnetic field was also studied for comparison. Products were characterized by scanning electron microscopy (SEM) and X-ray Diffraction (XRD). Results indicated that, regardless the precursor used, it was observed the presence of aluminum nanowires when the external magnetic mirror was applied, suggesting that magnetic fields are able to modify growth morphology at nanoscale

  9. Effects of modifying water environments on water supply and human health

    Science.gov (United States)

    Takizawa, S.; Nguyen, H. T.; Takeda, T.; Tran, N. T.

    2008-12-01

    Due to increasing population and per-capita water demand, demands for water are increasing in many parts of the world. Consequently, overuse of limited water resources leaves only small amounts of water in rivers and is bringing about rapid drawdown of groundwater tables. Water resources are affected by human activities such as excessive inputs of nutrients and other contaminants, agriculture and aquaculture expansions, and many development activities. The combined effects of modifying the water environments, both in terms of quantity and quality, on water supply and human health are presented in the paper with some examples from the Asian countries. In rural and sub-urban areas in Bangladesh and Vietnam, for example, the traditional way of obtaining surface water from ponds had been replaced by taking groundwaters to avert the microbial health risks that had arisen from contamination by human wastes. Such a change of water sources, however, has brought about human health impact caused by arsenic on a massive scale. In Thailand, the industrial development has driven the residents to get groundwater leaden with very high fluoride. Monitoring the urine fluoride levels reveal the risk of drinking fluoride-laden groundwaters. Rivers are also affected by extensive exploitation such as sand mining. As a result, turbidity changes abruptly after a heavy rainfall. In cities, due to shrinking water resources they have to take poor quality waters from contaminated sources. Algal blooms are seen in many reservoirs and lakes due to increasing levels of nutrients. Hence, it is likely that algal toxins may enter the water supply systems. Because most of the water treatment plants are not designed to remove those known and unknown contaminants, it is estimated that quite a large number of people are now under the threat of the public health "gtime bomb,"h which may one day bring about mass-scale health problems. In order to mitigate the negative impacts of modifying the water

  10. Measuring the pD of heavy water solutions

    International Nuclear Information System (INIS)

    Analogously to the chemistry of aqueous solutions, a chemistry of heavy water solutions can be developed and, correspondingly, by measuring the pD value - the activity of deuterium ions - the behavior of these solutions can be characterized. The cells for measurement of pD can be achieved similarly to the cell for pH, by using Ag/AgCl electrodes and thus a device for measuring the acid-basic balance in a heavy water solution can be made by measuring the electromotive force and establishing a corresponding pD scale. A strict comparison between pD and pH is not possible since both scales are conventional. Values of paD and paH for phosphate solutions of the same molar concentration in heavy and light water, respectively, differs constantly between them by 0.6 units. Such differences in various solutions, using glass electrodes and control calomel electrodes, allowed getting a more exact value (0.41 pH units) for solutions of molar concentration, with a standard deviation of 0.031. Thus, a pD value can be obtained with a pH-meter with glass and control calomel electrodes by applying the correction relation, plus 0,41 units. We have done such determinations by using a domestic made control electrode of the type HC Ag/AgCl pH=7. Determinations with sodium carbonate and iron and sodium chloride solutions in light and heavy water, at the same concentrations, were carried out. The values recorded from the pH-meter for heavy water solutions are lower than the pH values by 0.407 units, so that this value must be added to the pH readings to get the pD values. The standard deviation of these determinations was 0.012 pH units

  11. Structural analysis and modeling of water reactor fuel rod behavior

    International Nuclear Information System (INIS)

    An important aspect of the design and analysis of nuclear reactor is the ability to predict the behavior of fuel elements in the adverse environment of a reactor system under normal and emergency operating conditions. To achieve these objectives and in order to provide a suitable computer code based on fundamental material properties for design and study of the thermal-mechanical behavior of water reactor fuel rods during their irradiation life and also to demonstrate the fuel rod design and modeling for students, The KIANA-1 computer program has been developed by the writer at Amir-Kabir university of technology with support of Atomic Energy Organization of Iran. KIANA-1 is an integral one-dimensional computer program for the thermal and mechanical analysis in order to predict fuel rods performance and also parameter study of Zircaloy-clad UO2 fuel rod during steady state conditions. The code has been designed for the following main objectives: To give a solution for the steady state heat conduction equation for fuel as a heat source and clad by using finite difference, control volume and semi-analytical methods in order to predict the temperature profile in the fuel and cladding. To predict the inner gas pressures due to the filling gases and released gaseous fission products. To predict the fission gas production and release by using a simple diffusion model based on the Booth models and an empirical model. To calculate the fuel-clad gap conductance for cracked fuel with partial contact zones to a closed gap with strong contact. To predict the distribution of stress in three principal directions in the fuel and sheet by assuming one-dimensional plane strain and asymmetric idealization. To calculate the strain distribution in three principal directions and the corresponding deformation in the fuel and cladding. For this purpose the permanent strain such as creep or plasticity as well as the thermoelastic deformation and also the swelling, densification, cracking

  12. Study of plutonium recycling physics in light water reactors

    International Nuclear Information System (INIS)

    A stock of plutonium from the reprocessing of thermal neutron reactor fuel is likely to appear in the next few years. The use of this plutonium as fuel replacing 235U in thermal reactors is probably more interesting than simple stock-piling storage: immobilization of a capital which moreover would deteriorate by radioactive decay of isotope 241 also fissile and present to an appreciable extend in plutonium from reprocessing (half-life 15 years); recycling, on the other hand, will supply energy without complete degradation of the stock for fast neutron reactor loads, the burned matter having been partially renewed by conversion; furthermore the use of plutonium will meet the needs created by a temporary pressure on the naturel and/or enriched uranium market. For these two reasons the recycling of plutonium in thermal neutron reactors is being considered seriously today. The present work is confined to neutronic aspects and centres mainly on pressurized water-moderated reactors, the most highly developed at present in France. Four aspects of the problem are examined: 1. the physics of a plutonium-recycling reactor special features of neutronic phenomena with respect to the 'conventional' scheme of the 235U burning reactor; 2. calculation of a plutonium-recycling reactor: adaptation of standard methods; 3. qualification of these calculations from the viewpoint of both data and inevitable approximations; 4. the fuel cycle and particularly the equivalence of fissile matters

  13. Introduction of advanced pressurized water reactors in France

    International Nuclear Information System (INIS)

    Designed >30 yr ago, pressurized water reactors (PWRs) have evolved well to match the current safety, operating, and economic requirements. The first advanced PWR generation, the N4 reactor, is under construction with 1992 as a target date for commercial operation. The N4 may be considered to be a technological outcome of PWR evolution, providing advances in the fields of safety, man/machine interfaces, and load flexibility. As a step beyond N4, a second advanced PWR generation is presently under definition with, as a main objective, a greater ability to cope with the possible deterioration of the natural uranium market. In 1986, Electricite de France (EdF) launched investigations into the possible characteristics of this advanced PWR, called REP-2000 (PWR-2000: the reactor for the next century). Framatome joined EdF in 1987 but had been working on a new tight-lattice reactor. Main options are due by 1988; preliminary studies will begin and, by 1990, detailed design will proceed with the intent of firm commitments for the first unit by 1995. Commissioning is planned in the early years of the next century. This reactor type should be either an improved version of the N4 reactor or a spectral shift convertible reactor (RCVS). Through research and development efforts, Framatome, Commissariat a l'Energie Atomique (CEA), and EdF are investigating the physics of fuel rod tight lattices including neutronics, thermohydraulics, fuel behavior, and reactor mechanics

  14. Air quality impact analysis in support of the new production reactor environmental impact statement

    International Nuclear Information System (INIS)

    The Pacific Northwest Laboratory (PNL) conducted this air quality impact analysis for the US Department of Energy (DOE). The purpose of this work was to provide Argonne National Laboratory (ANL) with the required estimates of ground-level concentrations of five criteria air pollutants at the Hanford Site boundary from each of the stationary sources associated with the new production reactor (NPR) and its supporting facilities. The DOE proposes to provide new production capacity for the primary production of tritium and secondary production of plutonium to support the US nuclear weapons program. Three alternative reactor technologies are being considered by DOE: the light-water reactor, the low-temperature, heavy-water reactor, and the modular high-temperature, gas-cooled reactor. In this study, PNL provided estimates of the impacts of the proposed action on the ground-level concentration of the criteria air pollutants for each of the alternative technologies. The criteria pollutants were sulfur dioxide, nitrogen dioxide, carbon monoxide, total suspended particulates, and particulates with a diameter of less than 10 microns. Ground-level concentrations were estimated for the peak construction phase activities expected to occur in 1997 and for the operational phase activities beginning in the year 2000. Ground-level concentrations of the primary air pollutants were estimated to be well below any of the applicable national or state ambient air quality standards. 12 refs., 19 tabs

  15. 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

  16. The inner containment of an EPR trademark pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

    2014-08-15

    On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

  17. Standard- and extended-burnup PWR [pressurized-water reactor] and BWR [boiling-water reactor] reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    The purpose of this report is to describe an updated set of reactor models for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) operating on uranium fuel cycles and the methods used to generate the information for these models. Since new fuel cycle schemes and reactor core designs are introduced from time to time by reactor manufacturers and fuel vendors, an effort has been made to update these reactor models periodically and to expand the data bases used by the ORIGEN2 computer code. In addition, more sophisticated computational techniques than previously available were used to calculate the resulting reactor model cross-section libraries. The PWR models were based on a Westinghouse design, while the BWR models were based on a General Electric BWR/6 design. The specific reactor types considered in this report are as follows (see Glossary for the definition of these and other terms): (1) PWR-US, (2) PWR-UE, (3) BWR-US, (4) BWR-USO, and (5) BWR-UE. Each reactor model includes a unique data library that may be used to simulate the buildup and deletion of isotopes in nuclear materials using the ORIGEN2 computer code. 33 refs., 44 tabs

  18. Assessment of a small pressurized water reactor for industrial energy

    Energy Technology Data Exchange (ETDEWEB)

    Klepper, O. H.; Fuller, L. C.; Myers, M. L.

    1977-10-04

    An evaluation of several recent ERDA/ORNL sponsored studies on the application of a small, 365 MW(t) pressurized water reactor for industrial energy is presented. Preliminary studies have investigated technical and reliability requirements; costs for nuclear and fossil based steam were compared, including consideration of economic inflation and financing methods. For base-load industrial steam production, small reactors appear economically attractive relative to coal fired boilers that use coal priced at $30/ton.

  19. Natural uranium fueled light water moderated breeding hybrid power reactors

    International Nuclear Information System (INIS)

    The feasibility of fission-fusion hybrid reactors based on breeding light water thermal fission systems is investigated. The emphasis is on fuel-self-sufficient (FSS) hybrid power reactors that are fueled with natural uranium. Other LWHRs considered include FSS-LWHRs that are fueled with spent fuel from LWRs, and LWHRs which are to supplement LWRs to provide a tandem LWR-LWHR power economy that is fuel-self-sufficient

  20. Environmentally assisted cracking in light water reactors - annual report, January-December 2001.

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E; Hiller, R. W.; Shack, W. J.; Soppet, W. K.; Strain, R. V.; Energy Technology

    2003-06-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2001. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs, and (c) EAC of Alloy 600. The effects of key material and loading variables, such as strain amplitude, strain rate, temperature, dissolved oxygen (DO) level in water, and material heat treatment, on the fatigue lives of wrought and cast austenitic SSs in air and LWR environments have been evaluated. The mechanism of fatigue crack initiation in austenitic SSs in LWR environments has also been examined. The results indicate that the presence of a surface oxide film or difference in the characteristics of the oxide film has no effect on fatigue crack initiation in austenitic SSs in LWR environments. Slow-strain-rate tensile tests and post-test fractographic analyses were conducted on several model SS alloys irradiated to {approx}2 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) ({approx}3 dpa) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. Corrosion fatigue tests were conducted on nonirradiated austenitic SSs in high-purity water at 289 C to establish the test procedure and conditions that will be used for the tests on irradiated materials. A comprehensive irradiation experiment was initiated to obtain many tensile and disk specimens irradiated under simulated pressurized water reactor conditions at {approx}325 C to 5, 10, 20, and 40 dpa. Crack growth tests were completed on 30% cold-worked Alloy 600 in high-purity water under various environmental and loading conditions. The results are compared with data obtained earlier on several heats of Alloy 600

  1. Phenomenological modeling and study of a catalytic membrane reactor for water detritiation

    International Nuclear Information System (INIS)

    Tritium is produced in light and heavy water reactor fuel by ternary fission or neutron activation. This by-product is used as fuel in fusion fuel reactors such as JET in Culham or ITER in Cadarache (France). The growing interest of this research area will make the tritium fluxes increase; it is then worth addressing the question of its future whether it will be used or flushed out from liquid and gaseous effluents or waste. This thesis studies the recovery of tritium as fuel for fusion machines by means of packed bed membrane reactor (PBMR). Such a reactor combines catalytic conversion of tritiated water thanks to isotope exchange with hydrogen according to the reversible reaction Q2O+H2↔H2O+Q2 (Q=H,D or T) and selective permeation of Q2 through Pd-based membrane. In fact, palladium has the ability to bond with hydrogen isotopes, creating a selective permeation barrier. In the PBMR, thanks to the reaction products withdrawal, these permeation fluxes drive the heavy water conversion rate, to higher values than those reached in conventional fixed bed reactors (Le Chatelier's law). In order to study PBMRs, the CEA has built a test bench, using deuterium instead of tritium, allowing the analysis of their conversion and separation performances at the laboratory scale. An in-house method has been developed to determine simultaneously hydrogen and water isotopologues content by mass spectrometer analysis. It was experimentally shown that the activity of Ni-based catalyst used in this study was sufficient to allow the isotope exchange reactions to reach their thermodynamic equilibrium in a very short time. In addition, hydrogen permeation flux was shown to follow a Richardson's law. Sensitivity studies performed on the PBMR's main operating parameters revealed that its global performance (i.e. de-deuteration factor) increases with the temperature, the transmembrane pressure difference, the sweep gas flow rate and the residence time in the catalyst

  2. Nodal equivalence theory for hexagonal geometry, thermal reactor analysis

    International Nuclear Information System (INIS)

    An important aspect of advanced nodal methods is the determination of equivalent few-group parameters for the relatively large homogenized regions used in the nodal flux solution. The theoretical foundation for light water reactor (LWR) assembly homogenization methods has been clearly established, and during the last several years, its successes have secured its position in the stable of dependable LWR analysis methods. Groupwise discontinuity factors that correct for assembly homogenization errors are routinely generated along with the group constants during lattice physics analysis. During the last several years, there has been interest in applying equivalence theory to other reactor types and other geometries. A notable effort has been the work at Argonne National Laboratory to incorporate nodal equivalence theory (NET) for hexagonal lattices into the nodal diffusion option of the DIF3D code. This work was originally intended to improve the neutronics methods used for the analysis of the Experimental Breeder Reactor II (EBR-II), and Ref. 4 discusses the success of that application. More recently, however, attempts were made to apply NET to advanced, thermal reactor designs such as the modular high-temperature gas reactor (MHTGR) and the new production heavy water reactor (NPR/HWR). The same methods that were successful for EBR-II have encountered problems for these reactors. Our preliminary analysis indicates that the sharp global flux gradients in these cores requires large discontinuity factors (greater than 4 or 5) to reproduce the reference solution. This disrupts the convergence of the iterative methods used to solve for the node-wise flux moments and partial currents. Several attempts to remedy the problem have been made over the last few years, including bounding the discontinuity factors and providing improved initial guesses for the flux solution, but nothing has been satisfactory

  3. Gamma spectroscopy in water cooled reactors

    International Nuclear Information System (INIS)

    Gamma spectroscopy analysis of spent fuels in power reactors; study of two typical cases: determination of the power distribution by the mean of the activity of a low periodic element (Lanthanum 140) and determination of the burnup absolute rate by examining the ratio of Cesium 134 and Cesium 137 activities. Measures were realized on fuel solutions and on fuel assemblies. Development of a power distribution map of the assemblies and comparison with the results of a three dimensional calculation of core evolution

  4. Assessment of the thorium fuel cycle in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled.

  5. Removal and recovery of tritium from light and heavy water

    International Nuclear Information System (INIS)

    A method and apparatus for removing tritium from light water are described, comprising contacting tritiated feed water in a catalyst column in countercurrent flow with hydrogen gas originating from an electrolysis cell so as to enrich this feed water with tritium from the electrolytic hydrogen gas and passing the tritium enriched water to an electrolysis cell wherein the electrolytic hydrogen gas is generated and then fed upwards through the catalyst column or recovered as product. The tritium content of the hydrogen gas leaving the top of the enricher catalyst column is further reduced in a stripper column containing catalyst which transfers the tritium to a countercurrent flow of liquid water. Anodic oxygen and water vapour from the anode compartment may be fed to a drier and condensed electrolyte recycled with a slip stream or recovered as a further tritium product stream. A similar method involving heavy water is also described. (author)

  6. Solubility of carbohydrates in heavy water.

    Science.gov (United States)

    Cardoso, Marcus V C; Carvalho, Larissa V C; Sabadini, Edvaldo

    2012-05-15

    The solubility of several mono-(glucose and xylose), di-(sucrose and maltose), tri-(raffinose) and cyclic (α-cyclodextrin) saccharides in H(2)O and in D(2)O were measured over a range of temperatures. The solution enthalpies for the different carbohydrates in the two solvents were determined using the vant' Hoff equation and the values in D(2)O are presented here for the first time. Our findings indicate that the replacement of H(2)O by D(2)O remarkably decreases the solubilities of the less soluble carbohydrates, such as maltose, raffinose and α-cyclodextrin. On the other hand, the more soluble saccharides, glucose, xylose, and sucrose, are practically insensitive to the H/D replacement in water. PMID:22480785

  7. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    International Nuclear Information System (INIS)

    The technology of breeding 233U from 232Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program

  8. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    Energy Technology Data Exchange (ETDEWEB)

    Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

    1987-05-01

    The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.

  9. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Monado, Fiber [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su' ud, Zaki; Waris, Abdul; Basar, Khairul [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-09-30

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  10. Study on the adsorption of heavy metal ions from aqueous solution on modified SBA-15

    Directory of Open Access Journals (Sweden)

    Liliana Giraldo

    2013-01-01

    Full Text Available Amino-functionalized SBA-15 mesoporous silica was prepared, characterized, and used as an adsorbent for heavy metal ions. The organic - inorganic hybrid material was obtained by a grafting procedure using SBA-15 silica with 3-aminopropyl-triethoxysilane and bis(2,4,4-trimethylpentyl phosphinic acid (Cyanex 272, respectively. The structure and physicochemical properties of the materials were characterized by means of elemental analysis, X-ray diffraction (XRD, nitrogen adsorption - desorption, thermogravimetric analysis, FTIR spectroscopy and immersion calorimetry. The organic functional groups were successfully grafted onto the SBA-15 surface and the ordering of the support was not affected by the chemical modification. The behavior of the grafted solids was investigated for the adsorption of heavy metal ions from aqueous solutions. The hybrid materials showed high adsorption capacity and high selectivity for zinc ions. Other ions, such as cooper and cobalt were absorbed by the modified SBA-15 material.

  11. Fuel cladding interaction with water coolant in power reactors

    International Nuclear Information System (INIS)

    Water coolant chemistry and corrosion processes are important factors in reliable operation of NPP's, as at elevated temperatures water is aggressive towards structural materials. Water regimes for commercial Pressurized Water Reactors and Boiling Water Reactors were developed and proved to be satisfactory. Nevertheless, studies of operation experience continue and an amount of new Research and Development work is being conducted for further improvements of technology and better understanding of the physicochemical nature of those processes. In this report information is presented on the IAEA programme on fuel element cladding interaction with water coolant. Some results of this survey and recommendations made by the group of consultants who participated in this work are given as well as recommendations for continuation of this study. Separate abstracts were prepared for 6 papers of this report

  12. Design of virtual SCADA simulation system for pressurized water reactor

    International Nuclear Information System (INIS)

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor

  13. Design of virtual SCADA simulation system for pressurized water reactor

    Science.gov (United States)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  14. Novel Photocatalytic Reactor Development for Removal of Hydrocarbons from Water

    Directory of Open Access Journals (Sweden)

    Morgan Adams

    2008-01-01

    Full Text Available Hydrocarbons contamination of the marine environment generated by the offshore oil and gas industry is generated from a number of sources including oil contaminated drill cuttings and produced waters. The removal of hydrocarbons from both these sources is one of the most significant challenges facing this sector as it moves towards zero emissions. The application of a number of techniques which have been used to successfully destroy hydrocarbons in produced water and waste water effluents has previously been reported. This paper reports the application of semiconductor photocatalysis as a final polishing step for the removal of hydrocarbons from two waste effluent sources. Two reactor concepts were considered: a simple flat plate immobilised film unit, and a new rotating drum photocatalytic reactor. Both units proved to be effective in removing residual hydrocarbons from the effluent with the drum reactor reducing the hydrocarbon content by 90% under 10 minutes.

  15. Design of virtual SCADA simulation system for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

    2016-02-08

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  16. Methane production by treating vinasses from hydrous ethanol using a modified UASB reactor

    OpenAIRE

    España-Gamboa Elda I; Mijangos-Cortés Javier O; Hernández-Zárate Galdy; Maldonado Jorge A Domínguez; Alzate-Gaviria Liliana M

    2012-01-01

    Abstract Background A modified laboratory-scale upflow anaerobic sludge blanket (UASB) reactor was used to obtain methane by treating hydrous ethanol vinasse. Vinasses or stillage are waste materials with high organic loads, and a complex composition resulting from the process of alcohol distillation. They must initially be treated with anaerobic processes due to their high organic loads. Vinasses can be considered multipurpose waste for energy recovery and once treated they can be used in ag...

  17. 77 FR 23513 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-04-19

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft interim staff guidance; Request for public... Management Criteria for PWR Reactor Vessel Internal Components.'' The original notice provided the...

  18. Core surveillance of boiling-water reactors

    International Nuclear Information System (INIS)

    Methods suitable for a calculational procedure which determines the three-dimensional power distribution in boilingwater reactors on the basis of in-core detector readings are described. A two- dimensional equation based on diffusion theory is set up, and a method for incorporating detector readings in the solution of this equation is presented. A three-dimensional calculational method based on nodal theory is developed. Calculations are carried out using this method, and the results are compared with a three-dimensional nodal theory calculation . Finally, parameters affecting the detector readings are examined. (author)

  19. Reactivity events in Soviet-designed pressurized water reactors (VVERs)

    International Nuclear Information System (INIS)

    Analyses have been done on the response of Soviet-designed pressurized water reactors (VVERs) to various reactivity-induced events. VVERs are pressurized water reactors (PWRs), but they are different from US PWRs in many important ways. Significant differences exist in the design of the reactor protection system. Unlike US systems in which most off-normal conditions lead to a reactor scram, the Soviet system allows for four stages of response: full scram, fast insertion, slow insertion, and rod stop. For example, in a Soviet plant, the trip of a single reactor pump would not lead to a plant scram, but to the lowering of the reactor power to an appropriate level using the fast insertion response. The design of the Soviet system leads to additional failure modes not present in US plants. Reactivity transients in VVERs have been simulated using two different approaches: a simplified point kinetics model and the RETRAN-02 code. A total of 24 events were simulated for both VVER-440s and VVER-1000s. Several of these events, including deboration accidents at start-up and overcooling events without scram were found to lead to possible fuel damage

  20. Economics of advanced light water reactors - Recent update

    International Nuclear Information System (INIS)

    This paper includes recently updated analyses of the economic prospects of advanced light water reactors (ALWRs) during the decade of the 1990s. United Engineers and Constructors (UE and C) has performed engineering economic analyses related to ALWRs over the last 5 yr using both target economics and detailed cost-estimating methodologies. It has been found through such cost comparisons that properly designed and constructed ALWRs should cost less than the target cost figures listed above and significantly less than the pressurized water reactor better experience reference LWR plant cost

  1. Modified Dye for Water-Fast Ink-Jet Printing

    Institute of Scientific and Technical Information of China (English)

    Zheng Qingkang(郑庆康); Zhu Puxin(朱谱新); Wu Dacheng(吴大诚); Lewis DM.

    2001-01-01

    A quaternary ammonium compound containing amino acid residue was synthesized by Converting 3-chloro-2-hydroxy- propyalkyldimethylammonium chlorides into its epoxide derivatives, then attaching an amino acid to the epoxide derivatives synthesized a quaternary ammonium compound containing amino acid residue. Modified dyes were prepared by the ionotropy of anionic dyes with the quaternary ammonium compound containing amino acid residue. It was discovered that the modified dyes exhibited an excellent pH controllable solubility. These modified dyes have good water solubility at pH> 8.0, but they were water insoluble at pH < 6.5. On the printing paper, modified dyes in water-based ink-jet print ink could convert to water insoluble form and give prints excellent water fastness.

  2. Influence of Agriculture on Water Quality: Significance of Heavy Metals Monitoring

    OpenAIRE

    Nusreta Đonlagić; Amra Odobašić; Amra Bratovčić

    2007-01-01

    Agricultural activities directly influence the quality of water systems. Investigations showed that application of various agro-technical measures results with the pollution of water streams with heavy metals and other polluters. Increased concentrations of heavy metals result with intake of heavy metals and their transfer to food chains, and for that reason it is necessary to monitor the content of heavy metals regularly. Broad investigations of bio-geochemical cycling of heavy metals in the...

  3. Self-Sustaining Thorium Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud [Univ. of California, Berkeley, CA (United States); Gorman, Phillip M. [Univ. of California, Berkeley, CA (United States); Bogetic, Sandra [Univ. of California, Berkeley, CA (United States); Seifried, Jeffrey E. [Univ. of California, Berkeley, CA (United States); Zhang, Guanheng [Univ. of California, Berkeley, CA (United States); Varela, Christopher R. [Univ. of California, Berkeley, CA (United States); Fratoni, Massimiliano [Univ. of California, Berkeley, CA (United States); Vijic, Jasmina J. [Univ. of California, Berkeley, CA (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Hall, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Ward, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Jarrett, Michael [Univ. of Michigan, Ann Arbor, MI (United States); Wysocki, Aaron [Univ. of Michigan, Ann Arbor, MI (United States); Xu, Yunlin [Univ. of Michigan, Ann Arbor, MI (United States); Kazimi, Mujid [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Shirvan, Koroush [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mieloszyk, Alexander [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Todosow, Michael [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, Nicolas [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, Lap [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-03-15

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  4. Self-Sustaining Thorium Boiling Water Reactors

    International Nuclear Information System (INIS)

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  5. Recent developments in the target facilities at Argonne National Laboratory

    International Nuclear Information System (INIS)

    A description is given of recent developments in the target facility at Argonne National Laboratory. Highlights include equipment upgrades which enables us to provide enhanced capabilities for support of the Argonne Heavy-Ion ATLAS Accelerator Project. Also future plans and additional equipment acquisitions will be discussed. 3 refs., 3 tabs

  6. [Effect of heavy water on the viability of bacteria].

    Science.gov (United States)

    Dronova, N V; Parkhomenko, T V; Popov, V G; Sventitskiĭ, E N; Iakovleva, L Iu

    1988-01-01

    Influence of heavy water (D2O) on the membrane energization, the efflux of hydrogen ions and the respiration of bacteria E. coli M-17 was studied. As has been shown, heavy water of a low concentration (0.05-0.20% v/v) activates and of a high concentration (above 10%) inhibits the absorption of lipophilic cation tetraphenylphosphonium (TPP+) and of oxygen by cells. The return of these characteristics to the initial levels after the removal of D2O points to a reversible action of D2O. A protective effect of D2O towards membrane energization and rate of respiration on dried cells was observed. This fact is in agreement with the data on viability of bacteria. The indicated protective action increases at the stage of rehydration in the presence of D2O. PMID:3390482

  7. The temperature dependence of the rate constants and yields for the simulation of the radiolysis of heavy water

    International Nuclear Information System (INIS)

    At Chalk River Laboratories, a computer code is being developed to model the radiolysis of the heavy water in the moderator and the heat-transport system in CANDU reactors. This report collects together, for heavy water, the current knowledge regarding the rate constants, pKa's, yields and diffusion coefficients based on measurements in this laboratory and reports in the literature. The latest data available for the radiolysis of light water are generally included for comparison, which forms a partial update to the report on the radiolysis of light water (Elliot, AECL- 11073, COG-94-167, 1994). There are some reactions where little or no data are available at ambient or elevated temperatures; in these cases, an indication is given of the approach that will be taken to measure or estimate the required parameters. (author)

  8. Modified silicates applied in adsorption of heavy metal; Silicatos modificados aplicados na adsorcao de metal pesado

    Energy Technology Data Exchange (ETDEWEB)

    Farias, M.C.M. de [Universidade Federal de Campina Grande (UFCG), PB (Brazil). Unidade Academica de Engenharia de Materiais; Raposo, C.M.O., E-mail: raposo@dmg.ufcg.edu.b [Universidade Federal de Campina Grande (UFCG), Campina Grande, PB (Brazil). Unidade Academica de Mineracao e Geologia

    2010-07-01

    The levels of heavy metals in the environment has increased considerably in recent decades due to various human activities, which cause serious pollution problems, both in aquatic systems and in soil. The clay minerals present himself as amenable to the adsorption of metal ions and, sometimes, taking the advantage of being abundant and inexpensive. Vermiculite has intrinsic characteristics which favor its use as adsorbent. In this work, we investigate the adsorption of lead (II) from aqueous solutions by vermiculite fractions in commercial, fine to medium in molar concentration between 1-4 mmol (s). The samples provided by the Uniao Brasileira de Mineracao/Paraiba/Brazil were modified thermal and organically. The results of X-ray diffraction associated with the results of X-ray fluorescence showed that the average fraction vermiculite exfoliated organically modified responded most significantly to the adsorption process when compared to vermiculite fine fraction under the same conditions. (author)

  9. Integrated inspection programs at Bruce Heavy Water Plant

    International Nuclear Information System (INIS)

    Quality pressure boundary maintenance and an excellent loss prevention record at Bruce Heavy Water Plant are the results of the Material and Inspection Unit's five inspection programs. Experienced inspectors are responsible for the integrity of the pressure boundary in their own operating area. Inspectors are part of the Technical Section, and along with unit engineering staff, they provide technical input before, during, and after the job. How these programs are completed, and the results achieved, are discussed. 5 figs., 1 appendix

  10. Environmental health scoping study at Bruce Heavy Water Plant

    International Nuclear Information System (INIS)

    There are concerns that hydrogen sulfide released from the Heavy Water Plant near Kincardine, Ontario may be the cause of the mortalities and morbidities observed in a nearby flock of sheep. The Philosopher's Wool sheep farm is about four kilometres south-southeast of the Bruce Heavy Water Plant. Ontario Hydro, the owner and operator of the Bruce Heavy Water Plant, claims that hydrogen sulphide emissions from the Bruce Heavy Water Plant are within regulatory limits and well below levels that cause harm. Accordingly, the Atomic Energy Control Board commissioned the Alberta Environmental Centre, Alberta Department of Environmental Protection, to develop a scoping study for this environmental health issue. The first objective was to describe a field investigation model to define clearly the environmental health and operation of the sheep farm. The second objective was to describe possible exposure patterns and develop a holistic environmental pathway model. If appropriate, the third study objective was to describe animal models of the actual situation to elucidate specific aspects of the environmental health concerns. It was not the objective of this report to provide a definitive answer to the present environmental health issue. Ontario Hydro provided data to the Alberta Environmental Centre, as di the sheep farmer, the attending veterinarian, the University of Guelph study team, and the Atomic Energy Control Board. A six-tiered strategy of sequential evaluations of the ovine health problem is based on the multiple-response paradigm. It assumes the observed ovine health results are the result of multiple effector events. Each tier constitutes a separate, but inter-related, study. Sequential evaluation and feedback of each tier allow sound scientific judgements and efficient use of resources. (author). 59 refs., 11 tabs., 22 figs

  11. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  12. The Battle for Heavy Water Three physicists' heroic exploits

    CERN Multimedia

    2002-01-01

    Up until the end of the 1970s you could still catch a glimpse of his massive silhouette in the corridors of CERN. Lew Kowarksi, one of the pioneers of the Laboratory, was not only a great physicist; he was also a genuine hero of World War II. In 1940, along with Frédéric Joliot and Hans von Halban, Lew Kowarski managed to get the entire world supply of heavy water away to safety from the Nazis after a fantastic escape from occupied France. At the end of the war, the three physicists played themselves in a film about their adventures entitled 'la Bataille de l'eau lourde'. This film, which has been loaned to us by the French National Film Library, will be shown at CERN for the first time next Thursday. At the beginning of the war, heavy water (D20, two atoms of deuterium and one oxygen atom) was of strategic importance. In 1939 Frédéric Joliot, aided by Hans von Halban and Lew Kowarski, demonstrated the nuclear chain reaction and the moderator role that heavy water plays in it. A few weeks before the inv...

  13. DEGRADATION EVALUATION OF HEAVY WATER DRUMS AND TANKS

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J.; Vormelker, P.

    2009-07-31

    Heavy water with varying chemistries is currently being stored in over 6700 drums in L- and K-areas and in seven tanks in L-, K-, and C-areas. A detailed evaluation of the potential degradation of the drums and tanks, specific to their design and service conditions, has been performed to support the demonstration of their integrity throughout the desired storage period. The 55-gallon drums are of several designs with Type 304 stainless steel as the material of construction. The tanks have capacities ranging from 8000 to 45600 gallons and are made of Type 304 stainless steel. The drums and tanks were designed and fabricated to national regulations, codes and standards per procurement specifications for the Savannah River Site. The drums have had approximately 25 leakage failures over their 50+ years of use with the last drum failure occurring in 2003. The tanks have experienced no leaks to date. The failures in the drums have occurred principally near the bottom weld, which attaches the bottom to the drum sidewall. Failures have occurred by pitting, crevice and stress corrosion cracking and are attributable, in part, to the presence of chloride ions in the heavy water. Probable degradation mechanisms for the continued storage of heavy water were evaluated that could lead to future failures in the drum or tanks. This evaluation will be used to support establishment of an inspection plan which will include susceptible locations, methods, and frequencies for the drums and tanks to avoid future leakage failures.

  14. Plutonium breeding in liquid-metal fast breeder reactors and light water reactors

    International Nuclear Information System (INIS)

    The possibilities of breeding in liquid-metal fast breeder reactors (LMFBRs) and light water reactors (LWRs) are compared in two ways. The feasibility of breeding has been demonstrated in the Phenix reactor with a measured gain of 0.14. The gain in Superphenix will amount to about0.20. The studies show that while maintaining the performance of commercial reactors their breeding gain can be further increased either by the concept of heterogeneous cores or by using carbide or nitride fuel (breeding gain about0.35). Recently, the old idea of breeding in advanced pressurized water reactors (PWRs) has been taken up again with the objective of attaining a gain of 0.05. Unfortunately, these objectives had to be limited to a conversion ratio of 0.9 for safety reasons, and it is not certain whether operation will be rewarding economically. The strategy of substituting PWRs is examined using the French example. By gradually introducing LMFBRs, the cumulated uranium supplies in France can be kept within reasonable limits, which means that they attain three to four times the home resources. This is not possible with advanced LWRs, which can be considered only as a possible backup solution for plutonium recycling into PWRs

  15. Numerical study on seismic response of the reactor coolant pump in Advanced Passive Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    De, Cheng, E-mail: 0100209064@sjtu.edu.cn; Zhen-Qiang, Yao, E-mail: zqyaosjtu@gmail.com; Ya-bo, Xue; Hong, Shen

    2014-10-15

    Highlights: • An artificial accelerogram of the specified SSE is generated. • A dynamic FE model of the RCP in AP1000 (with gyroscopic and FSI effects) is developed. • The displacement, force, moment and stress in the RCP during the earthquake are summarized. - Abstract: The reactor coolant pump in the Advanced Passive Pressurized Water Reactor is a kind of nuclear canned-motor pump. The pump is classified as Seismic Category I, which must function normally during the Safe Shutdown Earthquake. When the nuclear power plant is located in seismically active region, the seismic response of the reactor coolant pump may become very important for the safety assessment of the whole nuclear power plant. In this article, an artificial accelerogram is generated. The response spectrum of the artificial accelerogram fits well with the design acceleration spectrum of the Safe Shutdown Earthquake. By applying the finite element modeling method, the dynamic finite element models of the rotor and stator in the reactor coolant pump are created separately. The rotor and stator are coupled by the journal bearings and the annular flow between the rotor and stator. Then the whole dynamic model of the reactor coolant pump is developed. Time domain analysis which uses the improved state-space Newmark method of a direct time integration scheme is carried out to investigate the response of the reactor coolant pump under the horizontal seismic load. The results show that the reactor coolant pump responds differently in the direction of the seismic load and in the perpendicular direction. During the Safe Shutdown Earthquake, the displacement response, the shear force, the moment and the journal bearing reaction forces in the reactor coolant pump are analyzed.

  16. Assessment of light water reactor accident management programs and experience

    International Nuclear Information System (INIS)

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation

  17. Assessment of light water reactor accident management programs and experience

    Energy Technology Data Exchange (ETDEWEB)

    Hammersley, R.J. [Fauske and Associates, Inc., Burr Ridge, IL (United States)

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation.

  18. Electrochemical sensors and devices for heavy metals assay in water: the French groups' contribution

    Science.gov (United States)

    Pujol, Luca; Evrard, David; Groenen-Serrano, Karine; Freyssinier, Mathilde; Ruffien-Ciszak, Audrey; Gros, Pierre

    2014-04-01

    A great challenge in the area of heavy metal trace detection is the development of electrochemical techniques and devices which are user-friendly, robust, selective, with low detection limits and allowing fast analyses. This review presents the major contribution of the French scientific academic community in the field of electrochemical sensors and electroanalytical methods within the last 20 years. From the well-known polarography to the up-to-date generation of functionalized interfaces, the different strategies dedicated to analytical performances improvement are exposed: stripping voltammetry, solid mercury-free electrode, ion selective sensor, carbon based materials, chemically modified electrodes, nano-structured surfaces. The paper particularly emphasizes their advantages and limits face to the last Water Frame Directive devoted to the Environmental Quality Standards for heavy metals. Recent trends on trace metal speciation as well as on automatic “on line” monitoring devices are also evoked.

  19. Electrochemical sensors and devices for heavy metals assay in water: the French groups' contribution

    Directory of Open Access Journals (Sweden)

    Luca ePUJOL

    2014-04-01

    Full Text Available A great challenge in the area of heavy metal trace detection is the development of electrochemical techniques and devices which are user-friendly, robust, selective, with low detection limits and allowing fast analyses. This review presents the major contribution of the French scientific academic community in the field of electrochemical sensors and electroanalytical methods within the last 20 years. From the well-known polarography to the up-to-date generation of functionalized interfaces, the different strategies dedicated to analytical performances improvement are exposed: stripping voltammetry, solid mercury-free electrode, ion selective sensor, carbon based materials, chemically modified electrodes, nano-structured surfaces. The paper particularly emphasizes their advantages and limits face to the last Water Frame Directive devoted to the Environmental Quality Standards for heavy metals. Recent trends on trace metal speciation as well as on automatic on line monitoring devices are also evoked.

  20. On the Effect of Lengthening Circadian Rhythm by Heavy Water

    Directory of Open Access Journals (Sweden)

    Akhmedov T. R.

    2014-01-01

    Full Text Available The problem of time sensor of biological clock (BC attracts interest of many scientists, and a great number of experiments are being conducted to stud y the influence of vari- ous physical and chemical factors on functioning of BC. Special attention is drawn to studying the influence of heavy water (D 2 O on functioning of BC that always leads to lengthening of circadian rhythms (CR. This work presents theoretical consideration of lengthening of CR, when hydrogen (H 2 in water is replaced by deuterium (D 2 , that is based on spacial difference of energy levels with similar principle quantum numbers.

  1. Research reactors

    International Nuclear Information System (INIS)

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  2. Isotopic signature of atmospheric xenon released from light water reactors

    International Nuclear Information System (INIS)

    A global monitoring system for atmospheric xenon radioactivity is being established as part of the International Monitoring System to verify compliance with the Comprehensive Nuclear-Test-Ban Treaty (CTBT). The isotopic activity ratios of 135Xe, 133mXe, 133Xe and 131mXe are of interest for distinguishing nuclear explosion sources from civilian releases. Simulations of light water reactor (LWR) fuel burn-up through three operational reactor power cycles are conducted to explore the possible xenon isotopic signature of nuclear reactor releases under different operational conditions. It is studied how ratio changes are related to various parameters including the neutron flux, uranium enrichment and fuel burn-up. Further, the impact of diffusion and mixing on the isotopic activity ratio variability are explored. The simulations are validated with reported reactor emissions. In addition, activity ratios are calculated for xenon isotopes released from nuclear explosions and these are compared to the reactor ratios in order to determine whether the discrimination of explosion releases from reactor effluents is possible based on isotopic activity ratios

  3. Evaluation of fatigue data including reactor water environmental effects

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [EPRI, Charlotte, NC (United States); Nickell, R.E. [Applied Science and Technology, Poway, CA (United States); Van Der Sluys, W.A. [Alliance, OH (United States); Yukawa, S. [Boulder, CO (United States)

    2002-07-01

    Laboratory data have been gathered in the past decade indicating a significant reduction in component fatigue life when reactor water environmental effects are experimentally simulated. However, these laboratory data have not been supported by nuclear power plant component operating experience. The laboratory data under simulated operating conditions are being used to support arguments for revising the design-basis fatigue curves in the ASME Code Section III, Division 1, for Class 1 components. A thorough review of available laboratory fatigue data and their applicability to actual component operating conditions was performed. The evaluation divided the assembly, review and assessment of existing laboratory fatigue data and its applicability to plant operating conditions into four principal tasks: (1) review of available laboratory data relative to thresholds for environmental parameters, such as temperature, reactor water oxidation potential, strain rate, strain amplitude, reactor water flow rate, and component metal sulfur content; (2) determination of the relevance of the laboratory data to actual plant operating conditions; (3) review of laboratory S-N data curve-fitting models; and (4) assessment of existing ASME Code Section III Class 1 margins This paper summarizes the results of the data review. In addition, recommendations are made for additional laboratory testing intended to improve the applicability of laboratory test results under simulated reactor water environmental conditions. (authors)

  4. European community light water reactor safety research projects. Experimental issue

    International Nuclear Information System (INIS)

    Research programs on light water reactor safety currently carried out in the European Community are presented. They cover: accident conditions (LOCA, ECCS, core meltdown, external influences, etc...), fault and accident prevention and means of mitigation, normal operation conditions, on and off site implications and equipment under severe accident conditions, and miscellaneous subjects

  5. Risk management and decision rules for light water reactor

    International Nuclear Information System (INIS)

    The process of developing and adopting safety objectives in quantitative terms can provide a basis for focusing societal decision making on the suitability of such objectives and upon questions of compliance with those objectives. A preliminary proposal for a light water reactor (LWR) risk management framework is presented as part of that process

  6. Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

    International Nuclear Information System (INIS)

    The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. Chemistry control strategies for a supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    The long-term viability of any Generation IV Supercritical Water-cooled Reactor (SCWR) concept depends on the ability of reactor designers and operators to predict and control water chemistry to minimize corrosion and corrosion product transport. Currently, SCWR material testing is being carried out using a limited range of water chemistries to establish the dependencies of the corrosion behavior on parameters such as water temperature and dissolved oxygen concentration. Once a final suite of candidate alloys is identified, more detailed, longer term testing will be needed under water chemistry regimes expected to be used in the SCWR. Prior to these tests, it will be necessary to identify proposed water chemistry regimes for the SCWR, and provide expected ranges for the key parameters. The direct-cycle design proposed for various SCWR concepts take advantage of the extensive operating experience world-wide of fossil-fired SCW power plants. Conceptually, the SCWR replaces the fossil-fired boiler with the reactor core (pressure vessel or pressure tube); the concept is broadly similar to that of a boiling water reactor. Current fossil-fired SCW power plants use either an all-volatile treatment or oxygenated water treatment for the feedwater systems. While the optimal water chemistry for a SCWR is yet to be determined, the monitored parameters are likely to be the same as those in existing fossil-fired and nuclear power plants: pH; conductivity, and concentrations of O2, H2, additives, impurities, corrosion and activation products. Monitoring will be required at many of the same components: feedwater, main 'steam', drains, pump outlets, condenser hotwell, and purification inlets and outlets. This paper outlines the strategy being used to develop a water chemistry regime for a CANDU® SCWR. It describes the key areas identified for chemistry monitoring and control: a) the reactor core, where materials are subjected to irradiation and high temperature, b

  10. New biosensor for detection of copper ions in water based on immobilized genetically modified yeast cells.

    Science.gov (United States)

    Vopálenská, Irena; Váchová, Libuše; Palková, Zdena

    2015-10-15

    Contamination of water by heavy metals represents a potential risk for both aquatic and terrestrial organisms, including humans. Heavy metals in water resources can come from various industrial activities, and drinking water can be ex-post contaminated by heavy metals such as Cu(2+) from house fittings (e.g., water reservoirs) and pipes. Here, we present a new copper biosensor capable of detecting copper ions at concentrations of 1-100 μM. This biosensor is based on cells of a specifically modified Saccharomyces cerevisiae strain immobilized in alginate beads. Depending on the concentration of copper, the biosensor beads change color from white, when copper is present in concentrations below the detection limit, to pink or red based on the increase in copper concentration. The biosensor was successfully tested in the determination of copper concentrations in real samples of water contaminated with copper ions. In contrast to analytical methods or other biosensors based on fluorescent proteins, the newly designed biosensor does not require specific equipment and allows the quick detection of copper in many parallel samples.

  11. New biosensor for detection of copper ions in water based on immobilized genetically modified yeast cells.

    Science.gov (United States)

    Vopálenská, Irena; Váchová, Libuše; Palková, Zdena

    2015-10-15

    Contamination of water by heavy metals represents a potential risk for both aquatic and terrestrial organisms, including humans. Heavy metals in water resources can come from various industrial activities, and drinking water can be ex-post contaminated by heavy metals such as Cu(2+) from house fittings (e.g., water reservoirs) and pipes. Here, we present a new copper biosensor capable of detecting copper ions at concentrations of 1-100 μM. This biosensor is based on cells of a specifically modified Saccharomyces cerevisiae strain immobilized in alginate beads. Depending on the concentration of copper, the biosensor beads change color from white, when copper is present in concentrations below the detection limit, to pink or red based on the increase in copper concentration. The biosensor was successfully tested in the determination of copper concentrations in real samples of water contaminated with copper ions. In contrast to analytical methods or other biosensors based on fluorescent proteins, the newly designed biosensor does not require specific equipment and allows the quick detection of copper in many parallel samples. PMID:25982723

  12. Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

  13. Safety aspects of designs for future light water reactors (evolutionary reactors)

    International Nuclear Information System (INIS)

    The main purpose of this document is to describe the major innovations of proposed designs of future light water reactors, to describe specific safety characteristics and safety analysis methodologies, and to give a general overview of the most important safety aspects related to future reactors. The reactors considered in this report are limited to those intended for fixed station electrical power production, excluding most revolutionary concepts. More in depth discussion is devoted to those designs that are in a more advanced state of completion and have been more extensively described and analysed in the open literature. Other designs will be briefly described, as evidence of the large spectrum of new proposals. Some designs are similar; others implement unique features and require specific discussion (not all aspects of designs with unique features are fully discussed in this document). 131 refs, 22 figs

  14. Post-remedial-action survey report for Kinetic Experiment Water Boiler Reactor Facility, Santa Susana Field Laboratories, Rockwell International, Ventura County, California

    International Nuclear Information System (INIS)

    Rockwell International's Santa Susana Laboratories in Ventura County, California, have been the site of numerous federally-funded contracted projects involving the use of radioactive materials. Among these was the Kinetics Experiment Water Boiler (KEWB) Reactor which was operated under the auspices of the US Atomic Energy Commission (AEC). The KEWB Reactor was last operated in 1966. The facility was subsequently declared excess and decontamination and decommissioning operations were conducted during the first half of calendar year 1975. The facility was completely dismantled and the site graded to blend with the surrounding terrain. During October 1981, a post-remedial-action (certification) survey of the KEWB site was conducted on the behalf of the US Department of Energy by the Radiological Survey Group (RSG) of the Occupational Health and Safety Division's Health Physics Section (OHS/HP) of Argonne National Laboratory (ANL). The survey confirmed that the site was free from contamination and could be released for unrestricted use

  15. Fatigue and environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Fatigue and stress corrosion cracking (SCC) for low-alloy steel used in piping and in steam generator and reactor pressure vessels have been investigated. Fatigue data were obtained on medium-sulfur-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor water, and in air. Analytical studies focused on the behavior of carbon steels in boiling water reactor (BWR) environments. Crack-growth rates of composite fracture-mechanics specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B steel were determined under small-amplitude cyclic loading in HP water with ∼300 pbb dissolved oxygen. Radiation-induced segregation and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence also have been investigated. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain-rate tensile tests were conducted on tubular specimens in air and in simulated BWR water at 289 degrees C

  16. Environmentally assisted cracking in light water reactors : semiannual report, July 2000 - December 2000.

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.; Energy Technology

    2002-04-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from July 2000 to December 2000. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. The fatigue strain-vs.-life data are summarized for the effects of various material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Effects of the reactor coolant environment on the mechanism of fatigue crack initiation are discussed. Two methods for incorporating the effects of LWR coolant environments into the ASME Code fatigue evaluations are presented. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. A fracture toughness J-R curve test was conducted on a commercial heat of Type 304 SS that was irradiated to {approx}2.0 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. The results were compared with the data obtained earlier on steels irradiated to 0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) (0.45 and 1.35 dpa). Neutron irradiation at 288 C was found to decrease the fracture toughness of austenitic SSs. Tests were conducted on compact-tension specimens of Alloy 600 under cyclic loading to evaluate the enhancement of crack growth rates in LWR environments. Then, the existing fatigue crack growth data on Alloys 600 and 690 were analyzed to establish the effects of temperature, load ratio, frequency, and stress intensity range

  17. Elementary Analyses and Heavy Metal Contents of Tap Waters in Konuralp District: Comparison of Mains Water, Spring Water and Zamzam

    Directory of Open Access Journals (Sweden)

    Muammer Yılmaz 1

    2014-09-01

    Full Text Available Objective: We meet our water needs such as city water supply, natural spring water, Zamzam water was aimed to compare in terms of chemical ion concentration and heavy metal content. Methods: City water from the four regions with different source, Zamzam water and bottled natural spring water in samples, ions and heavy metal values measured. Results have been assessed according to the criteria specified in the United States environmental protection agency (EPA and the World Health Organization (WHO. Results: In the sample of tap water taken from Konuralp, Al and Fe values were found over the EPA-WHO limit value. In the sample of bottled natural spring water, heavy metals are within the limits established. In the sample of Zamzam water Ca+2, Mg+2 values were higher than other samples but not exceeding the limits. In the sample of Zamzam water nitrate (NO3-, and vanadium (V values is very high from samples taken of the city water. Conclusion: Water content may be different with the water supply and environmental effects. More extensive analysis should be done by municipalities to drinking water that contains ions and heavy metal and citizens to know the measurements of the water they drink should be informed periodically of local authority’s websites.

  18. Study on soluble heavy metals with preconcentration by using a new modified oligosilsesquioxane sorbent

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Eduardo G.; Soares, Isaac V. [Departamento de Fisica e Quimica, Unesp-Univ Estadual Paulista, Av. Brasil, 56-Centro, Caixa Postal 31, 15385-000 Ilha Solteira, SP (Brazil); Filho, Newton L. Dias, E-mail: nldias@dfq.feis.unesp.br [Departamento de Fisica e Quimica, Unesp-Univ Estadual Paulista, Av. Brasil, 56-Centro, Caixa Postal 31, 15385-000 Ilha Solteira, SP (Brazil); Silva, Nileia C. da; Perujo, Sergio D.; Bastos, Andrea C.; Garcia, Edemir F.; Ferreira, Tamires T. [Departamento de Fisica e Quimica, Unesp-Univ Estadual Paulista, Av. Brasil, 56-Centro, Caixa Postal 31, 15385-000 Ilha Solteira, SP (Brazil); Fraceto, Leonardo F.; Rosa, Andre H. [Departamento de Engenharia Ambiental, Unesp-Univ Estadual Paulista, Avenida Tres de Marco, n Degree-Sign 511, CEP 18087-180 Sorocaba, SP (Brazil)

    2012-10-30

    Highlights: Black-Right-Pointing-Pointer A modified oligosilsesquioxane sorbent was evaluated for its heavy metal adsorption. Black-Right-Pointing-Pointer The adsorption capacities are much higher than those presented by other adsorbents. Black-Right-Pointing-Pointer The method was successfully applied to the analysis of environmental samples. - Abstract: In this work, a new modified oligosilsesquioxane was prepared, characterized and evaluated for its heavy metal adsorption characteristics from aqueous solution. The material was characterized using infrared spectroscopy, X-ray diffractometry, solid-state {sup 29}Si and {sup 13}C nuclear magnetic resonance. Batch and column experiments were conducted to investigate for Fe (III), Cr (III), Cu (II), Cd (II), Pb (II) and Ni (II) removal from dilute aqueous solution by sorption onto modified oligosilsesquioxane. The results obtained in the flow experiments, showed a recovery of ca. 100% of the metal ions adsorbed in a column packed with 2.0 g of nanomaterial, using 3 mL of 1.0 mol L{sup -1} HCl solution as eluent. The detection limits for Fe, Cr, Cu, Cd, Pb, and Ni were 0.26, 0.33, 0.38, 0.41, 0.47 and 0.55 {mu}g L{sup -1}, respectively. The relative standard deviation values (n = 12) were <4.82% for Fe, <4.12% for Cr, <3.79% for Cu, <3.03% for Cd, <4.02% for Pb and <2.65% for Ni. The proposed method was successfully applied to the analysis of environmental samples.

  19. Stainless steel clad for light water reactor fuels. Final report

    International Nuclear Information System (INIS)

    Proper reactor operation and design guidelines are necessary to assure fuel integrity. The occurrence of fuel rod failures for operation in compliance with existing guidelines suggests the need for more adequate or applicable operation/design criteria. The intent of this study is to develop such criteria for light water reactor fuel rods with stainless steel clad and to indicate the nature of uncertainties in its development. The performance areas investigated herein are: long term creepdown and fuel swelling effects on clad dimensional changes and on proximity to clad failure; and short term clad failure possibilities during up-power ramps

  20. Domestic light water reactor fuel design evolution. Volume III

    International Nuclear Information System (INIS)

    Volume III of this report examines the design evolution of domestic light water reactor fuel. The fuel of each vendor is individually described. Tables and figures detail the fuel's design parameters. A data base of this nature is required for the design of an underwater fuel disassembly and rod storage system. An assessment of fuel failure mechanisms and fuel performance is presented showing that spent fuel pool operational problems will be minimal or nonexistent. A summary and projection of spent fuel discharges, organized by reactor and fuel design type, is included to show the magnitude and composition of the spent fuel situation facing the nuclear industry

  1. Cooling water treatment for heavy water project (Paper No. 6.9)

    International Nuclear Information System (INIS)

    With minor exceptions, water is the preferred industrial medium for the removal of unwanted heat from process systems. The application of various chemical treatments is required to protect the system from water related and process related problems of corrosion, scale and deposition and biofouling. The paper discusses the cooling water problems for heavy water industries along with the impact caused by associated fertilizer units. (author). 6 figs

  2. BUCKLING ANALYSES OF A HEAVY COLUMN CONSIDERATED IN WATER

    Directory of Open Access Journals (Sweden)

    Yeliz PEKBEY

    2008-02-01

    Full Text Available In 1744, the critical buckling load with the assumption of uniform cross-section without weight of column were computed by Euler. Whenever an economical solution is required, the weight of column must be considered for solution of buckling analyses. In literature, the critical buckling load and asymptotic behaviour of heavy column in condition of atmosphere have inverstigated for ten different support types. When this literature is examined, it is stated that the differential equations of for four different suppport types in condition of water is similar to condition of atmosphere. However, the differential equations of other four different suppport types in condition of water is different from to condition of atmosphere. And it is stated that the critical buckling load these different suppport types in condition of water is not calculated from condition of atmosphere. The goals of this paper are to develop self weight buckling of column at its top fixed and lower end fixed-roller supported in condition of water. This paper, presents a analytical method for calculating the critical buckling load of the heavy column.

  3. Improvements in water reactor fuel technology and utilization

    International Nuclear Information System (INIS)

    The International Symposium on Improvements in Water Reactor Fuel Technology and Utilization was organized by the International Atomic Energy Agency and held in Stockholm from 15 to 19 September 1986 at the invitation of the Government of Sweden. The aim was to give scientists and engineers working in these fields the opportunity to exchange information on their achievements to date and their future work. It was attended by about 170 participants from 29 Member States and one international organization. A total of 37 papers and 12 posters covering a wide range of topics related to water reactor fuel was presented. The number of participants as well as the large number of fuel vendors from Europe, Japan and the United States of America and some reactor utilities proved that the timing and the topic of the Symposium were well chosen. In general, the Symposium has shown that current water reactor fuels perform reliably and meet current performance requirements. The factors which could limit fuel performance under high burnup conditions and load follow mode of operation were discussed and defined. All 49 presentations were divided into 6 sections: introduction (3 general papers); fuel design and performance (15 papers); fuel materials and behavior (8 papers); structural materials (5 papers); fuel fabrication (6 papers) and poster section (12 papers). A separate abstract was prepared for each of these presentations

  4. Computational Fluid Dynamics Analysis of Canadian Supercritical Water Reactor (SCWR)

    Science.gov (United States)

    Movassat, Mohammad; Bailey, Joanne; Yetisir, Metin

    2015-11-01

    A Computational Fluid Dynamics (CFD) simulation was performed on the proposed design for the Canadian SuperCritical Water Reactor (SCWR). The proposed Canadian SCWR is a 1200 MW(e) supercritical light-water cooled nuclear reactor with pressurized fuel channels. The reactor concept uses an inlet plenum that all fuel channels are attached to and an outlet header nested inside the inlet plenum. The coolant enters the inlet plenum at 350 C and exits the outlet header at 625 C. The operating pressure is approximately 26 MPa. The high pressure and high temperature outlet conditions result in a higher electric conversion efficiency as compared to existing light water reactors. In this work, CFD simulations were performed to model fluid flow and heat transfer in the inlet plenum, outlet header, and various parts of the fuel assembly. The ANSYS Fluent solver was used for simulations. Results showed that mass flow rate distribution in fuel channels varies radially and the inner channels achieve higher outlet temperatures. At the outlet header, zones with rotational flow were formed as the fluid from 336 fuel channels merged. Results also suggested that insulation of the outlet header should be considered to reduce the thermal stresses caused by the large temperature gradients.

  5. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Science.gov (United States)

    2011-10-03

    ... Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... published in the Federal Register on October 21, 2010, (75 FR 65038-65039). Detailed meeting agendas...

  6. Trend of inspection and evaluation, repair and preventive maintenance guidelines for light water reactor internals

    International Nuclear Information System (INIS)

    Inspection and Evaluation Guidelines for Light Water Reactor Internals have been enacted because the damages of stress corrosion cracking etc. were found recently in light water reactor internals. This paper describes the trend of the guidelines. (author)

  7. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.''...

  8. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR)...

  9. Core design calculations of IRIS reactor using modified CORD-2 code package

    International Nuclear Information System (INIS)

    Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)

  10. Study on core design for reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  11. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  12. COMSORS: A light water reactor chemical core catcher

    International Nuclear Information System (INIS)

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B2O3) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation

  13. Reactivity Impact of 2H and 16O Elastic Scattering Nuclear Data on Critical Systems with Heavy Water

    Science.gov (United States)

    Roubtsov, D.; Kozier, K. S.; Chow, J. C.; Plompen, A. J. M.; Kopecky, S.; Svenne, J. P.; Canton, L.

    2014-04-01

    The accuracy of deuterium nuclear data is important for reactor physics simulations of heavy water (D2O) reactors. The elastic neutron scattering cross section data at thermal energies, σs,th, have been observed to have noticeable impact on the reactivity values in simulations of critical systems involving D2O. We discuss how the uncertainties in the thermal scattering cross sections of 2H(n,n)2H and 16O(n,n)16O propagate to the uncertainty of the calculated neutron multiplication factor, keff, in thermal critical assemblies with heavy water neutron moderator/reflector. The method of trial evaluated nuclear data files, in which specific cross sections are individually perturbed, is used to calculate the sensitivity coefficients of keff to the microscopic nuclear data, such as σs(E) characterized by σs,th. Large reactivity differences of up to ≃ 5-10 mk (500-1000 pcm) were observed using 2H and 16O data files with different elastic scattering data in MCNP5 simulations of the LANL HEU heavy-water solution thermal critical experiments included in the ICSBEP handbook.

  14. INFLUENCE OF HEAVY METALS HYDROXIDES ON WATER DISSOCIATION IN BIPOLAR MEMBRANE

    Directory of Open Access Journals (Sweden)

    Sheldeshov N. V.

    2015-12-01

    Full Text Available The results of study of bipolar membrane – analogue of MB-2, modified with chemically introduced chromium (III, iron (III and nickel (II hydroxides by the method of frequency spectrum of electrochemical impedance, by infrared spectroscopy and scanning electron microscopy in combination with X-ray spectrum analysis are presented. It is shown, that sequential treatment of cation-exchanger, contained in cationexchange membrane, with metal salt solution and alkali solution does not result in formation of complex compounds of these metals with ionic groups of ion exchanger. It was found that in these conditions the presence of heavy metals in the phase of cationexchanger confirmed by X-ray analysis, however, crystals of hydroxides of heavy metals are not detected in the size range of 1000 nm to 20 nm. These heavy metal compounds are thermally unstable and their catalytic activity in the reaction of dissociation of water molecules decreases with increasing temperature during heat treatment. The introduction of low-soluble hydroxides of d-metals (chromium (III, iron (III, nickel(II by chemical method can significantly improve the electrochemical characteristics of a bipolar membrane. The most effective catalysts in water dissociation reaction are the hydroxides of chromium (III and iron (III and, as a consequence, membranes with these hydroxides have a lower value of overpotential compared with original membrane at the same current density

  15. Hydrodynamics and Mass Transfer in a Modified Three-phase Airlift Loop Reactor

    Institute of Scientific and Technical Information of China (English)

    Liu Mengxi; Lu Chunxi; Shi Mingxian; Ge Baoli; Huang Jie

    2007-01-01

    A modified internal-loop airlift reactor (MIALR) with a continuous slurry phase was studied to investigate the local hydrodynamic characteristics, including gas holdup, bubble size, bubble rise velocity and local mass transfer properties. Based on the analysis of geometrical construction and fluid properties of gas and slurry, MIALR was divided into six flow regions. In these flow regions, the local hydrodynamic characteristics were investigated over a wide range of operating variables. Furthermore, a new method was developed to measure the dissolved oxygen concentration. The volumetric mass-transfer coefficient in six flow regions was also calculated for comparison.

  16. Deposition of heavy water on soil and reemission to the atmosphere

    International Nuclear Information System (INIS)

    Field experiments using heavy water as a tracer instead of tritiated water (HTO) were carried out in November 1995 and August 1996 in Japan. The objective of these experiments was to estimate the behavior of HTO in the environment when HTO was released to the atmosphere. We measured the evolution of depth profiles of heavy water concentrations in soil water and compared the reemission rates with the evaporation velocities to study the deposition and reemission of heavy water to/from soil. The depth profiles of heavy water concentrations in soil were expressed by exponential functions of which the gradient depended on the deposition period. The initial reemission rates of heavy water from the soil were the highest, regardless of the meteorological conditions. The reemission occurred not only during the day but also at night, suggesting that the exchange of heavy water in the soil surface layer with H2O in air played a dominant role during nighttime. (orig.)

  17. Water chemistry management in cooling system of research reactor in JAERI

    International Nuclear Information System (INIS)

    The department of research reactor presently operates three research reactors (JRR-2, JRR-3M and JRR-4). For controlling and management of water and gas in each research reactor are performed by the staffs of the research reactor technology development division. Water chemistry management of each research reactor is one of the important subject. The main objects are to prevent the corrosion of water cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the radioactive waste. In this report describe a outline of each research reactor facilities, radiochemical analytical methods and chemical analytical methods for water chemistry management. (author)

  18. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-01-20

    ... COMMISSION 10 CFR Part 52 RIN 3150-AI84 U.S. Advanced Boiling Water Reactor Aircraft Impact Design... the U.S. Advanced Boiling Water Reactor (ABWR) standard plant design to comply with the NRC's aircraft...--Design Certification Rule for the U.S. Advanced Boiling Water Reactor IV. Section-by-Section Analysis...

  19. 78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

    Science.gov (United States)

    2013-10-25

    ... COMMISSION Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors AGENCY... Core Cooling Systems for Pressurized-Water Reactors.'' This RG is being revised to incorporate guidance... emergency core cooling systems (ECCSs) of pressurized water reactors (PWRs). This RG also describes...

  20. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs)....