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Sample records for argonne heavy water modified reactor

  1. Heavy water reactors physics

    International Nuclear Information System (INIS)

    An important research programme on heavy water reactor physics has been carried out in France for quite a few years. The decision to build the EL 4 prototype and so to choose the heavy water gas cooled type has renewed the interest in this programme and at the same time given to it a more specific orientation A summary of the results gained in this field is presented in this paper. In the first part are described the experimental investigations, most of them were carried out in the criticality facility AQUILON II. The experiments are grouped in four parts - Systematic studies of lattices Buckling measurements. - Specific studies of gas-cooled lattices. - Fine structure, spectral indices measurements etc... - Measurements on lattices or samples containing Uranium of various enrichment or Plutonium. The second part is devoted to a summary of the theoretical studies. The whole results have allowed an improvement of the calculation methods, have led to a better understanding of the neutron balance in lattices, and have permitted the establishment of a set of formula to predict not only the clean fuel conditions but also the evolution of the nuclear properties with irradiation. Some specific studies on power reactor are quoted. (authors)

  2. 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

  3. 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)

  4. 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)

  5. Fast neutron flux in heavy water reactors

    International Nuclear Information System (INIS)

    The possibility of calculating the fast neutron flux in a natural uranium-heavy water lattice by superposition of the individual contributions of the different fuel elements was verified using a one-dimension Monte-Carlo code. The results obtained are in good agreement with experimental measurements done in the core and reflector of the reactor AQUILON. (author)

  6. 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.)

  7. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors)

  8. Heavy Water Components Test Reactor Decommissioning

    International Nuclear Information System (INIS)

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D and D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment

  9. General description of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor is a boiling light water cooled, heavy water moderated and vertical pressure tube type reactor with its design optimised for utilisation of thorium for power generation. The core consists of (Th-U233 )O2 and (Th-Pu)O2 fuel with a discharge burn up of 20,000 MWd/Te. This reactor incorporates several features to simplify the design, which eliminate certain systems and components. AHWR design is also optimised for easy replaceability of coolant channels, facilitation of in-service inspection and maintenance and ease of erection. The AHWR design also incorporates several passive systems for performing safety-related functions in the event of an accident. In case of LOCA, emergency coolant is injected through 4 accumulators of 260 m3 capacity directly into the core. Gravity driven water pool of capacity 6000 m3 serves to cool the core for 3 days without operator's intervention. Core submergence, passive containment isolation and passive containment cooling are the added features in AHWR. The paper describes the various process systems, core and fuel design, primary components and safety concepts of AHWR. Plant layout and technical data are also presented. The conceptual design of the reactor has been completed, and the detailed design and development is scheduled for completion in the year 2002. (author)

  10. Outlook for Heavy-Water Reactors

    International Nuclear Information System (INIS)

    For large capacity units heavy-water reactors have already established an economic position competitive with any other nuclear or conventional electric generators in certain quite normal circumstances. With continued development they are expected to maintain this position and contribute a large fraction of the world's power demand into the indefinite future and extend to other large-scale uses such as desalting water. The grounds for this forecast are shown to be firmly based on a low-fuelling cost despite any likely rise in the price of uranium, together with extreme flexibility to adapt the fuel cycle to meet any changes in the market of fuel supply and reprocessing. Appropriate use of thorium with uranium is likely in the long term. As the scale of use expands, the cost of fabricating fuel and of producing heavy water will fall. Very few large heavy-water power reactors have yet been built and as experience is gained the cost of construction will fall. An important near-term development will be the practice of taking heat from the fuel direct to generate the working steam. In the longer term as improved materials are developed for higher temperatures the thermal to electrical conversion efficiency will be raised. With all these improvements in view, a long-term competitive position is foreseen. Details are given of several fuel cycle changes showing the inventories of fuel and heavy water involved in meeting market changes and the expectation of net fuel cycle costs being held in the range 0.3 to 0.6 mill/kWh in terms of to-day's dollar values despite a rise in the cost of uranium. The analysis shows possible interaction with other types of reactor including breeders. Reasons are given for welcoming any possible future competition or partnership from these types in meeting the growing world demand for harnessed energy. (author)

  11. 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

  12. Water chemistry features of advanced heavy water reactor

    International Nuclear Information System (INIS)

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

  13. Loss of feed water analyses of advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. Passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power level with no primary coolant pumps. The case analysed in this paper is the loss of feedwater to steam drum which results in decrease in heat removal from core. This also causes increase in reactor pressure. Further consequences depend upon various protective and engineered safeguard systems like relief system, reactor trip, isolation condenser and advanced accumulator. Analysis has been done using code RELAP5/MOD3.2. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, qualities and flow in different part of Primary Heat Transport (PHT) system. (author)

  14. Behaviour of heavy water in nuclear reactors of the CEA

    International Nuclear Information System (INIS)

    In the two heavy water reactors of the CEA: Zoe and P-2, we do: A) the supervision of the isotopic composition of the heavy water; B) the supervision of gases released by the decomposition of the heavy water under radiation, and to their recombination; C) periodic analyses of impurities. (M.B.)

  15. 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)

  16. 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

  17. Project plan for the decontamination and decommissioning of the Argonne National Laboratory Experimental Boiling Water Reactor

    International Nuclear Information System (INIS)

    In 1956, the Experimental Boiling Water Reactor (EBWR) Facility was first operated at Argonne National Laboratory (ANL) as a test reactor to demonstrate the feasibility of operating an integrated power plant using a direct cycle boiling water reactor as a heat source. In 1967, ANL permanently shut down the EBWR and placed it in dry lay-up. This project plan presents the schedule and organization for the decontamination and decommissioning of the EBWR Facility which will allow it to be reused by other ANL scientific research programs. The project total estimated cost is $14.3M and is projected to generate 22,000 cubic feet of low-level radioactive waste which will be disposed of at an approved DOE burial ground. 18 figs., 3 tabs

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

    International Nuclear Information System (INIS)

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

  19. 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.

  20. 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

  1. Decontamination of the heavy water system of the RA reactor

    International Nuclear Information System (INIS)

    The heavy water system of the RA reactor was decontaminated of 60Co. The solution used for decontamination was 7% H3PO4 and 3% CrO3. The decontamination factor ranged from 10 to 100. From the results the distribution of 60Co in the heavy water, and on stainless and aluminium parts was determined (author)

  2. Analysis of Gamma Activity of Heavy Water at RB Reactor

    International Nuclear Information System (INIS)

    The RB experimental nuclear reactor still works with heavy water obtained in 1959 from the former USSR. Gamma activity of the heavy water was periodically controlled during the past time. In this experiment, measurements were carried out with two samples: D2O taken from the RB reactor and D2O that has been used in the reactor. Two germanium spectrometers were used as detectors. Gamma spectra data were evaluated manually and using several computer codes. results of the experiment show that gamma activity D2O of RB reactor is at the level of background in the Vinca Institute, without contamination with fission products. (author)

  3. 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)

  4. Possibility of using metal uranium fuel in heavy water reactors

    International Nuclear Information System (INIS)

    The review of metal uranium properties including irradiation in the reactor core lead to the following conclusions. Using metal uranium in the heavy water reactors would be favourable from economic point of view for ita high density, i.e. high conversion factor and low cost of fuel elements fabrication. Most important constraint is swelling during burnup and corrosion

  5. Detection of gaseous heavy water leakage points in CANDU 6 pressurized heavy water reactors

    International Nuclear Information System (INIS)

    During reactor operation, the heavy water filled primary coolant system in a CANDU 6 Pressurized Heavy Water (PHWR) may leak through routine operations of the plant via components, mechanical joints, and during inadvertent operations etc. Early detection of leak points is therefore important to maintain plant safety and economy. There are many independent systems to monitor and recover heavy water leakage in a CANDU 6 PHWR. Methodology for early detection based on operating experience from these systems, is investigated in this paper. In addition, the four symptoms of D2O leakage, the associated process for clarifying and verifying the leakage, and the probable points of leakage are discussed. (author)

  6. 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

  7. Possibilities of using metal uranium fuel in heavy water reactors

    International Nuclear Information System (INIS)

    There are serious economic reasons for using metal uranium in heavy water reactors, because of its high density, i.e. high conversion factor, and low cost of fuel elements production. Most important disadvantages are swelling at high burnup and corrosion risk. Some design concepts and application of improved uranium obtained by alloying are promising for achievement of satisfactory stability of metal uranium under reactor operation conditions

  8. Study of a Heavy-Water Reactor with Boiling Heavy-Water Coolant

    International Nuclear Information System (INIS)

    Among the possible types of heavy-water reactor, those cooled by heavy water would appear to combine the advantages of excellent neutron economy and a well-tried cladding material; this allows optimum utilization of uranium under the present conditions of technology. Placing the reactor, the handling equipment, and the heat exchangers together in a prestressed concrete vessel appreciably simplifies operating problems by reducing the number of hermetic seals in contact with the pressurized heavy water. This arrangement is only effective if a large proportion of the heat transfer is by phase change, so as to keep the amount of coolant to a minimum. The Commissariat à Energie Atomique has made a study of a boiling heavy-water reactor under a co-operation agreement with the Siemens and Sulzer Companies and with the participation of the Socia Company. The paper describes the main features of these projects as well as the main technological problems raised by this design which relate to the thermal insulation of the concrete vessel in the presence of a two-phase fluid; the handling equipment which must function in steam at 300°C; and the accessibility of the exchangers. (author)

  9. Decontamination and decommissioning of the Experimental Boiling Water Reactor at Argonne National Laboratory

    International Nuclear Information System (INIS)

    The Experimental Boiling Water Reactor (EBWR), located on the Argonne National Laboratory-East (ANL-E) site, started operations in 1957. The initial rating was 20 MW(t). The rating was eventually increased to 70 MW(t) in 1959 and 100 MW(t) in 1962. The reactor was shut down in 1967 and all of the fuel was removed from the facility. The facility was placed in dry lay-up until 1986. ANL-E personnel started the decontamination and decommissioning (D ampersand D) effort in 1986. Supporting equipment such as the external steam system and some of the upper reactor components, the core riser and the top fuel shroud, were removed at that time. Characterization of the facility was also undertaken. The contract to complete the EBWR D ampersand D Project was issued in December 1993. The initial schedule called for the final effort to be divided into five phases that were to be completed over a four year period. However, this schedule was subsequently consolidated, at the request of ANL-E, to a thirteen month period, with the on-site work to be completed by the end of 1994. The EBWR D ampersand D Project is approximately 88% complete. A small quantity of reactor internals remains to be volume reduced along with the removal of the SFSP water treatment system. Upon completion of this work the facility will be decontaminated and a final survey completed. The planned completion of on-site work is scheduled for July 1995

  10. 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

  11. 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.

  12. 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)

  13. 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...

  14. 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...

  15. 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.

  16. 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)

  17. Fuel for CANDU pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Unique properties, performance and evolution of CANDU fuel are described. The manufacturing conditions, uranium requirements, and fuel costs are discussed. The in-service performance of the fuel has been excellent and defect mechanisms and operating criterion are described. Evolutionary improvements in CANDU fuel and new fuel cycles such as plutonium and thorium are being explored to insure that the CANDU reactor remains competitive in the future. (author)

  18. 20% inlet header break analysis of Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. A passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power levels, with no primary coolant pumps. Loss of coolant due to failure of inlet header results in depressurization of primary heat transport (PHT) system and containment pressure rise. Depressurization activates various protective and engineered safety systems like reactor trip, isolation condenser and advanced accumulator, limiting the consequences of the event. This paper discusses the thermal hydraulic transient analysis for evaluating the safety of the reactor, following 20% inlet header break using RELAP5/MOD3.2. For the analysis, the system is discretized appropriately to simulate possible flow reversal in one of the core paths during the transient. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, steam quality and flow in different parts of the Primary Heat Transport (PHT) system. Flow and energy discharges into the containment are also estimated for use in containment analysis. (author)

  19. Advanced reactor design and safety objectives - The heavy water reactor perspective

    International Nuclear Information System (INIS)

    This paper provides a summary of the major requirements for future nuclear reactors from CANDU operating station owners based on the various studies and plans prepared. Most of the specific technical requirements for Advanced Heavy Water reactor Systems are based on systematic reviews of current operating CANDU stations to identify opportunities for generic improvements in reliability, operability, maintainability and to address emerging licensing or safety issues. Hence these requirements represent those for the evolutionary development of the Advanced Heavy Water Reactor systems factoring in the considerable operating experience of the CANDU stations. This evolutionary approach to the development of advanced heavy water reactors will be consistent with a philosophy of minimizing the risk to future reactor owners whose requirements are for a reliable, low cost unit

  20. 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

  1. Advanced heavy water reactor pressure tube-easy replaceability

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor (AHWR) is a 300 MWe vertical pressure tube type reactor. A coolant channel consists of pressure tube, made of Zr-2.5 % Nb, which is separated from cold calandria tube using garter spring spacers. The principal function of pressure tube is to support and locate the fuel assembly and allows light water coolant through fuel assembly by natural circulation. Since AHWR is designed for life of 100 years, it necessitates the replacement of pressure tubes during service life. Easy replaceability of pressure tube, along with surveillance requirements, has major bearing on the design of coolant channel assembly. The several systems and tools have been conceptualised to cater the needs for easy and quick replacement of a pressure tube during reactor shut down. This paper gives the highlights of the innovative design features of coolant channel, preliminary design and pre-requisites for replacement, and experimental programme for demonstration of easy replaceability. (author)

  2. 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...

  3. Planning for the decommissioning of a heavy water research reactor

    International Nuclear Information System (INIS)

    The Heavy Water Research Reactor (HWRR) was constructed and put into operation in 1958 at the China Institute of Atomic Energy (CIAE), located in the suburbs of Beijing. It was the first nuclear reactor in China. The HWRR is a 10 MW multipurpose research reactor and has been operated for 48 years. Because of its long operating history and aged equipment, it is scheduled to be finally shut down by the end of 2007. It has been decided by CIAE to implement a strategy of immediate dismantling after final shutdown. The paper describes the preparation work for the development of the HWRR decommissioning plan at CIAE. The establishment and organization of the project and the problems encountered are described. Progress and problems are addressed. The paper also discusses the measures needed for the successful planning of decommissioning. (author)

  4. Accident analysis of heavy water cooled thorium breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yulianti, Yanti [Department of Physics, University of Lampung Jl. Sumantri Brojonegoro No.1 Bandar Lampung, Indonesia Email: y-yanti@unila.ac.id (Indonesia); Su’ud, Zaki [Department of Physics, Bandung Institute of Technology Jl. Ganesha 10 Bandung, Indonesia Email: szaki@fi.itb.ac.id (Indonesia); Takaki, Naoyuki [Department of Nuclear Safety Engineering Cooperative Major in Nuclear Energy (Graduate School) 1-28-1 Tamazutsumi,Setagayaku, Tokyo158-8557, Japan Email: ntakaki@tcu.ac.jp (Japan)

    2015-04-16

    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 < 0.2% dk/k, and negative coolant reactivity coefficient. One of the 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

  5. Accident analysis of heavy water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    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 < 0.2% dk/k, and negative coolant reactivity coefficient. One of the 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

  6. 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 coolant reactivity coefficient. One of the 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

  7. On the physics design of Advanced Heavy Water Reactor (AHWR)

    International Nuclear Information System (INIS)

    Full text: The AHWR is a 920 MWth, vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water. The prime objective is to produce power utilizing thorium available abundantly in India from a relatively simple system with enhanced safety level. It is endowed with several innovative safety features such as negative coolant void reactivity, heat removal through natural circulation and passive containment cooling. The development of reactor design has drawn heavily on the experience generated through design and operation of Pressurised Heavy Water Reactors (PHWR) and Boiling Water Reactor (BWR) in India. It was an opportunity to develop a reactor system using thorium-based fuel and gain some valuable experience. A non-proliferative thorium/U-233 based closed fuel cycle is chosen for AHWR. Plutonium discharged from PHWRs is used as the fissile seed fuel with thorium for the generation of U-233 and then as a top-up fuel in the equilibrium core along with self-sustaining U-233 in the thorium matrix. The physics design has several challenges in achieving negative void reactivity, spatial core control, on-line fuelling and minimization of inventory of plutonium fuel. It is difficult to achieve negative coolant void coefficient in a heavy water moderated pressure tube type reactor. For this a multi-pronged approach involving pitch reduction, heterogeneous cluster design and use of mild absorbers is chosen. Plutonium bearing fuel is located separately in the outer region of the cluster with self-sustaining U-233 bearing fuel in the inner region of the cluster. A small amount of mild absorber is located in the centre of the cluster. The void coefficient varies with burnup and it is a challenge to have it negative throughout the core. The state of nuclear data for the elements of interest and type of neutron spectrum in the reactor puts heavy demand on the calculation models and validation of reactivity coefficients to ensure

  8. Argonne heavy ion fusion program

    International Nuclear Information System (INIS)

    The experimental part of Argonne's heavy ion fusion program is directed toward demonstrating the first, and in many ways most difficult, section of a viable accelerator facility for heavy ion fusion. this includes a high current, high brightness, singly charged xenon source, a dc preaccelerator at the highest practical voltage, and a low beta linac of special design. The latter would demonstrate rf capture with its attendant inefficiencies and accelerate ions to a velocity acceptable to more conventional rf linac structures such as the π-3π Wideroe. The initial goals of this program are for a source current of 100 mA of Xe+1, a preaccelerator voltage of 1.5 MV, and less than 50% loss in rf capture into the low beta linac. A linear accelerator is proposed with a voltage gain up to 200 MV as a minimum which would form the initial stage of an operational heavy ion fusion facility irrespective of what type of acceleration to high energies were employed beyond this point

  9. The development of reactor vessel internal heavy forging for 1000 MW pressurized-water reactor nuclear power plant

    International Nuclear Information System (INIS)

    This Paper introduced the development of Reactor Vessel Internal (RVI) heavy forgings for 1000 MW Pressurized Water Reactor (PWR) nuclear power plant, analyzed the manufacture difficulties and technical countermeasures. The testing result of the product indicated that the performance of RVI heavy forgings manufactured by Shanghai Heavy Machinery Plant Ld. (SHMP) is outstanding and entirely satisfy the technical requirements for RVI product. (authors)

  10. The effect of heavy water reactors and liquid fuel reactors on the long-term development of nuclear energy

    International Nuclear Information System (INIS)

    The effects of the rates at which various combinations of power reactor types are installed on the long-range (to the year 2040) uranium and plutonium inventory requirements are examined. Consideration is given to light water reactors, fast breeder reactors, high temperature gas-cooled reactors, heavy water reactors, and thermal breeder reactors, in various combinations, and assuming alternatively a 3% and a 5% growth in energy demand

  11. Design improvements for pressurized heavy water reactors in Korea

    International Nuclear Information System (INIS)

    Three medium size (700 MWe class) CANDU (CANadian Deuterium Uranium) 6 Pressurized Heavy Water Reactors (PHWR) are currently under construction at Wolsong site in Korea by KEPCO (Korea Electric Power Corporation). KAERI (Korea Atomic Energy Research Institute) and AECL (Atomic Energy of Canada Limited) have jointly undertaken studies on incorporating improved design features and technology based on the operating plant experience. The improvements for future PHWRs in Korea was described and compared to the current CANDU 6 and other operating CANDU units. The CANDU 9 PHWR based on the single unit adaptation of Darlington plant operating in Canada with natural uranium fuel utilizes proven technology throughout. The reactor design has 480 fuel channels compared with 380 fuel channels for the CANDU 6 reactor. CANDU 9 design features can be further updated so that the Improved PHWR will meet evolving utility requirements in Korea. The Improved PHWR with a higher electrical output than CANDU 6 offers a number of advantages to utilities with compatible grid capacity. These include economy of scale in capital cost and operating cost. Future Improved PHWRs can further achieve a higher output by using recovered uranium or slightly enriched uranium. (author)

  12. 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.)

  13. Cooling of concrete structure in advanced heavy water reactor

    International Nuclear Information System (INIS)

    Innovative nuclear power plants are being designed by incorporation of passive systems to the extent possible for enhancing the safety by elimination of active components. BARC has designed Advanced Heavy Water Reactor (AHWR) incorporating several passive systems to facilitate the fulfillment of safety functions of the reactor during normal operation, residual heat removal, emergency core cooling, confinement of radioactivity etc. In addition to these passive systems, an innovative passive technology is being developed to protect, the concrete structure in high temperature zone (V1-volume). Passive Concrete Cooling System (PConCS) uses the principle of natural circulation to provide cooling outside the insulation cabinet encompassing high temperature piping. Cooling water is circulated from overhead GDWP in cooling pipes fixed over corrugated plate on outer surface of insulation cabinet and maintains low temperature of concrete structure. Modular construction of insulation cabinet and cooling pipes external to the concrete surface simplifies the design, construction and refurbishment if required. The paper describes the details of passive technology for concrete cooling. (author)

  14. Physics design of advanced heavy water reactor utilising thorium

    International Nuclear Information System (INIS)

    An Advanced Heavy Water Reactor (AHWR) is being developed in India with the aim of utilising thorium for power generation. AHWR is a vertical pressure tube type reactor cooled by boiling light water and moderated by heavy water. It has been optimised for the thorium cycle. The main design objective is to be self-sustaining in 233U with most of the power from the thorium fuel using plutonium as the external fissile feed. It incorporates several advanced safety features namely, heat removal through natural circulation and a negative void coefficient of reactivity. The reactor has been designed to produce 750 MW(th) at a discharge burnup of 20,000 MWd/H(e). The physics design of AHWR has followed an evolutionary path ranging from a seed and blanket concept to a simplified composite cluster to achieve a good thermal hydraulic coupling. We have designed a composite cluster using both kinds of fuel namely, (Th-UO2 and (Th-Pu)O2. With plutonium seed, negative void coefficient can be achieved by making the spectrum harder. This was done by using a pyrocarbon scatterer in the moderator. The void coefficient strongly depends on plutonium. As plutonium burns very rapidly, it is not possible to achieve uniformly negative void coefficient with burnup in this cluster. Alternatively, burnable poison can be used within the cluster to achieve negative void coefficient taking advantage of the flux redistribution and change in spectrum upon voiding. Here, it is possible to achieve almost constant void reactivity with burnup resulting in a good thermal hydraulic coupling. The cluster design presently incorporates a central burnable absorber region. Boiling light water coolant requires that the core power distribution be optimised with thermal hydraulic parameters. The peaking factors inside the cluster should be low so as to have significant margin in operational conditions and to avoid burnout in accident conditions. The variation of reactivity from cold clean to hot operating has

  15. 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

  16. 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.

  17. 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)

  18. 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

  19. Advanced reactor design and safety objectives. The heavy water reactor perspective

    International Nuclear Information System (INIS)

    The development objectives of advanced heavy water reactors (AHWRs) should be guided by the requirements of the operating utilities. The paper provides a summary of the major requirements for future nuclear reactors from CANDU operating station owners based on the various studies and plans prepared. Most of the specific technical requirements for AHWR systems are based on systematic reviews of current operating CANDU stations. Hence these requirements represent those for the evolutionary development of AHWR systems, factoring in the considerable operating experiences of the CANDU stations. The requirements for the new HWR designs can be summarized under economic objectives, safety objectives, operational objectives and other utility requirements. 2 refs

  20. 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.

  1. 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

  2. Toxicity of irradiated advanced heavy water reactor fuels.

    Science.gov (United States)

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels. PMID:23274823

  3. Construction management of Indian pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Pandit Jawaharlal Nehru and Dr. Homi J. Bhabha, the visionary architects of Science and Technology of modern India foresaw the imperative need to establish a firm base for indigenous research and development in the field of nuclear electricity generation. The initial phase has primarily focused on the technology development in a systematic and structured manner, which has resulted in establishment of strong engineering, manufacturing and construction base. The nuclear power program started with the setting up of two units of boiling light water type reactors in 1969 for speedy establishment of nuclear technology, safety culture, and development of operation and maintenance manpower. The main aim at that stage was to demonstrate (to ourselves, and indeed to the rest of the world) that India, inspite of being a developing country, with limited industrial infrastructure and low capacity power grids, could successfully assimilate the high technology involved in the safe and economical operation of nuclear power reactors. The selection of a BWR was in contrast to the pressurized heavy water reactors (PHWR), which was identified as the flagship for the first stage of India's nuclear power program. The long-term program in three stages utilizes large reserves of thorium in the monazite sands of Kerala beaches in the third stage with first stage comprising of series of PHWR type plants with a base of 10,000 MW. India has at present 14 reactors in operation 12 of these being of PHWR type. The performance of operating units of 2720 MW has improved significantly with an overall capacity factor of about 90% in recent times. The construction work on eight reactor units with installed capacity of 3960 MW (two PHWRs of 540 MW each, four PHWRs of 220 MW each and two VVERs of 1000 MW each) is proceeding on a rapid pace with project schedules of less than 5 years from first pour of concrete. This is being achieved through advanced construction technology and management. Present

  4. Investigation of the heavy water distillation system at the RA reactor

    International Nuclear Information System (INIS)

    The heavy water distillation system was tested because this was not done before the reactor start-up. Detailed inspection of the system components showed satisfactory results. Leak testing was done as well as the testing of the instrumentation which enables reliable performance of the system. Performance testing was done with ordinary water and later 2700 l of heavy water from the reactor was purified, decreasing the activity by 45%

  5. 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)

  6. Reactor D and D at Argonne National Laboratory - lessons learned

    International Nuclear Information System (INIS)

    This paper focuses on the lessons learned during the decontamination and decommissioning (D and D) of two reactors at Argonne National Laboratory-East (ANL-E). The Experimental Boiling Water Reactor (EBWR) was a 100 MW(t), 5 MSV(e) proof-of-concept facility. The Janus Reactor was a 200 kW(t) reactor located at the Biological Irradiation Facility and was used to study the effects of neutron radiation on animals

  7. Determination of the tritium content in the reactor heavy water, Phase II

    International Nuclear Information System (INIS)

    Measurement results of the 3H activity in non-irradiated water and after reactor operation are presented. Methods were developed for sampling and radiochemical water purification by ion exchange and multiple distillation. Methods for absolute measurement of soft beta radiation of tritium were established. Migration of tritium through the heavy water RA reactor system was monitored. Results were compared with other measured reactor parameters

  8. 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

  9. 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.

  10. 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.

  11. 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

  12. 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

  13. 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.

  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. 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

  16. Halden Boiling Water Reactor. Plant Performance and Heavy-Water Management

    International Nuclear Information System (INIS)

    The Halden boiling heavy-water reactor, designed and built by the Norwegian Institutt for Atomenergi, has since June 1958 been operated as an international project. On its second charge the reactor was operated at power levels up to 25 MW and most of the time at a pressure of 28.5 kg/cm2. During the period from July 1964 to December 1966 the plant availability was close to 64% including shutdowns because of test fuel failures and loading/unloading of fuel. Disregarding such stops, the availability was close to 90%. The average burnup of the core is about 6200 MWd/t UO2 : the most highly exposed elements have reached 10000 MWd/t UO2. The transition temperature of the reactor tank has been followed closely. The results of the surveillance programme and the implication on the reactor operation are discussed. The reactor is located in a cave in a rock. Some experiences with such a containment are given. To locate failed test-fuel elements a fuel failure location system has been installed. A fission gas collection system has saved valuable reactor time during clean-up of the reactor system following test fuel failures. Apart from one incident with two of the control stations, the plant control and instrumentation systems have functioned satisfactorily. Two incidents with losses of 150 and 200 kg of heavy water have occurred. However, after improved methods for leakage detection had been developed, the losses have been kept better than 50 g/h . Since April 1962 the isotopic purity of the heavy water (14 t) has decreased from 99.75 to 99.62%. The tritium concentration is now slightly above 700 μC/cm3. This activity level has not created any serious operational or maintenance problems. An extensive series of water chemistry experiments has been performed to study the influence of various operating parameters on radiolytic gas formation. The main results of these experiments will be reported. Different materials such as mild steel, ferritic steel and aluminium have been

  17. Emergency electric power supply systems for Pressurised Heavy Water Reactor

    International Nuclear Information System (INIS)

    This safety guide is one of a series of guides, already prepared or are under preparation as a follow-up of the Code of Practice on Design for Safety in Pressurised Heavy Water Based Nuclear Power Plants. The guide is based on the current design of 220 MWe and 500 MWe PHWRs. This safety guide specifically provides guidance on all aspects of safety in designing an emergency electric power supply system and about the basic requirements of other types of power supply systems in NPPs

  18. 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.

  19. 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

  20. 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.

  1. 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)

  2. Analysis of severe accidents in pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Certain very low probability plant states that are beyond design basis accident conditions and which may arise owing to multiple failures of safety systems leading to significant core degradation may jeopardize the integrity of many or all the barriers to the release of radioactive material. Such event sequences are called severe accidents. It is required in the IAEA Safety Requirements publication on Safety of the Nuclear Power Plants: Design, that consideration be given to severe accident sequences, using a combination of engineering judgement and probabilistic methods, to determine those sequences for which reasonably practicable preventive or mitigatory measures can be identified. Acceptable measures need not involve the application of conservative engineering practices used in setting and evaluating design basis accidents, but rather should be based on realistic or best estimate assumptions, methods and analytical criteria. Recently, the IAEA developed a Safety Report on Approaches and Tools for Severe Accident Analysis. This publication provides a description of factors important to severe accident analysis, an overview of severe accident phenomena and the current status in their modelling, categorization of available computer codes, and differences in approaches for various applications of severe accident analysis. The report covers both the in- and ex-vessel phases of severe accidents. The publication is consistent with the IAEA Safety Report on Accident Analysis for Nuclear Power Plants and can be considered as a complementary report specifically devoted to the analysis of severe accidents. Although the report does not explicitly differentiate among various reactor types, it has been written essentially on the basis of available knowledge and databases developed for light water reactors. Therefore its application is mostly oriented towards PWRs and BWRs and, to a more limited extent, they can be only used as preliminary guidance for other types of reactors

  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. Feasibility study and economic analysis on thorium utilization in heavy water reactors

    International Nuclear Information System (INIS)

    Even though natural uranium is a more easily usable fuel in heavy water reactors, thorium fuel cycles have also been considered owing to certain attractive features of the thorium fuel cycle in heavy water reactors. The relatively higher fission neutron yield per thermal neutron absorption in 233U combined with the very low neutron absorption cross section of heavy water make it possible to achieve breeding in a heavy water reactor operating on Th-233U fuel cycle. Even if the breeding ratio is very low, once a self-sustaining cycle is achieved, thereafter dependence on uranium can be completely eliminated. Thus, with a self-sustaining Th-233U fuel cycle in heavy water reactors, a given quantity of natural uranium will be capable of supporting a much larger installed generating capacity to significantly longer period of time. However, since thorium does not contain any fissile isotope, fissile material has to be added at the beginning. Concentrated fissile material is considerably more expensive than the 235U contained in natural uranium. This makes the fuel cycle cost higher with thorium fuel cycle, at least during the initial stages. The situation is made worse by the fact that, because of its higher thermal neutron absorption cross section, thorium requires a higher concentration of fissile material than 238U. Nevertheless, because of the superior nuclear characteristics of 233U, once uranium becomes more expensive, thorium fuel cycle in heavy water reactors may become economically acceptable. Furthermore, the energy that can be made available from a given quantity of uranium is considerably increased with a self-sustaining thorium fuel cycle

  5. 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

  6. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    International Nuclear Information System (INIS)

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR

  7. Health physics problems due to tritium around heavy-water reactors

    International Nuclear Information System (INIS)

    A review is made, after one year operation, of the protection against tritium in a heavy-water reactor: - nature of the tritium risk, particular to this type of installation; - detection and measurement equipment; - working method; - results obtained concerning the supervision of the personnel, the installation, and the servicing operations. (authors)

  8. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1996-08-01

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR.

  9. Neutronic study of the two french heavy water reactors

    International Nuclear Information System (INIS)

    The two french reactors - the reactor of Chatillon, named Zoe, and the reactor of Saclay - P2 - were the object of detailed neutronic studies which the main ideas are exposed in this report. These studies were mostly done by the Department of the Reactor Studies (D.E.P.). We have thus studied the distribution of neutronic fluxes; the factors influencing reactivity; the link between reactivity and divergence with the formula of Nordheim; the mean time life of neutrons; neutron spectra s of P2; the xenon effect; or the effect of the different adjustments of the plates and controls bar. (M.B.)

  10. Designing a plant for tritium and protium removal from heavy water of the PIK reactor

    International Nuclear Information System (INIS)

    The project of the facility for tritium and protium isotopic removal (FIR) from the PIK reactor heavy water is described. The FIR principle scheme, wherein different methods for isotopes separation and management of deuterium, contaminated by tritium are applied, is presented. The experimental facilities created for obtaining the initial data required for the FIR project development are described. In particular, the facility for the hydrogen isotopes separation is created on the basis of the isotope exchange between water and hydrogen and water electrolysis

  11. Design guide for category II reactors light and heavy water cooled reactors

    International Nuclear Information System (INIS)

    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

  12. 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.

  13. Problems related with the power regulation of reactors by physico-chemical methods, and the behaviour of water and heavy water in nuclear reactors

    International Nuclear Information System (INIS)

    The observation of the behaviour of water and heavy water in working reactors contributes to their safe running and provides information useful in studies relating to reactivity control techniques using soluble poisons. The use of nuclear poisons dissolved in the water of a reactor leads to its chemical pollution. The conditions under which they can be used without causing the undesirable effects of this pollution are studied. Problems of analysis, although important, are not tackled in this paper. Behaviour of heavy water in working reactors. Isotopic pollution of heavy water: The rate at which this pollution occurs depends on the type of reactor and on certain characteristic incidents. The use of a re-concentration column is an efficient way of maintaining the heavy water isotopic concentration in a reactor which cannot be considered exempt from slow isotopic pollution. Heavy water leak detection: The instantaneous rates of small leaks can be measured, the leak localised, and the atmospheric contamination in the reactor building checked. Isotopic analysis for deuterium or tritium determination, is carried out on samples. Chemical pollution and purification of heavy water Chemical pollution of heavy water is one of the most complex problems in reactor chemistry Corrosion of the materials which make up the core and the heavy water circuits varies extensively with the state of purity of the heavy water, as may be appreciated from the performances of the purification circuits and from direct measurements From the knowledge acquired it has been possible to work out standards of purity which, if observed will guarantee satisfactory running of the reactor. Radiolytic decomposition of heavy water : A better knowledge of its quantitative aspects in reactors is necessary in order to foresee the amounts of explosive gas mixture given off in power reactors. The radiolysis rate develops with the chemical purity of the water and the instantaneous power of the reactor

  14. 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.

  15. Thermal hydraulic analysis due to the changes in heat removal for advanced heavy water reactor

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor is a natural circulation light water cooled and heavy water moderated pressure tube reactor. Changes in heat removal by primary heat transport system of the reactor have significant impact on various important system parameters like pressures, qualities, reactor power and flows. Increase in heat removal leads to the cooldown of the system subsequently reducing pressure, void increase and changes in power and flows of the system. Decrease in heat removal leads to warm-up of the system subsequently raising pressure, void collapse, and changes in power and flows of the system. The behaviour is complex as system under consideration is natural circulation system. This article presents the results of simulations made with the RELAP5-MOD3.2 code that show first the impact of a decrease in feed water temperature on fluid temperature, steam drum pressure, core exit void, reactivity, reactor power, core flow, steam flow and clad temperature and secondly the impact of a loss of normal feed water flow on steam drum pressure, channel flow, core quality, clad surface temperature. For lowering of feed water temperature transient and in isolation condensers cold water injection, the reactor power increases and the reactor trips on the high power signal. Simultaneous flow increment due to the 2 phase natural circulation characteristic has caused the clad temperature to limit to their steady state value. In case of loss of feed transient the reactor trips on high pressure. The clad surface temperature rise from steady state operating value is marginal and it is well within the safety limit as per the acceptance criteria

  16. International conference on Future of Heavy Water Reactors (HWR-FUTURE)

    International Nuclear Information System (INIS)

    The International Conference on Future of Heavy Water Reactors (HWR-FUTURE) was held in Ottawa, Ontario, Canada on October 2-5, 2011. The conference comprised of various case studies, presentations and roundtable discussions on several pertinent topics dealing with the fast emerging Heavy Water Reactor or HWR technology. This conference addressed the challenges faced by the sector and assisted in planning feasible methods to combat them. Emphasizing effectively on developments and issues, knowledge exchange and technology transfer, and establishing prospective collaborations on reactor design, fuel design, material and chemistry, thermal hydraulics and safety, and operating experience for HWRs, presented were paper presentations, plenary sessions, live demos and practical solutions pertaining to these issues.

  17. 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

  18. 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

  19. Chemical elimination of alumina in suspension in nuclear reactors heavy water

    International Nuclear Information System (INIS)

    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)

  20. Observations on the removal of gadolinium from the moderator system of pressurised heavy water reactor (PHWR) and advanced heavy water reactor (AHWR)

    International Nuclear Information System (INIS)

    Investigation on ion exchange removal of gadolinium taken as gadolinium nitrate, which is used as neutron poison in the moderator system of Pressurised Heavy Water Reactor (PHWR) and proposed to be used in Advanced Heavy Water Reactor (AHWR) was carried out. Mixed bed operation consisting of (a) strong acid cation resin (SAC) and strong base anion resin (SBA) and (b) strong acid action resin and acrylic acid based nitrate loaded weak base anion resin were employed for the removal gadolinium from its aqueous solution at pH 5. In the former case, the outlet of the mixed bed was highly alkaline, which resulted in precipitation of gadolinium hydroxide. In the latter case, the pH of the system never crossed 6 and gadolinium was effectively picked up on the resin without getting precipitated. Series operation consisting for strong acid cation resin followed by mixed bed column consisting of strong acid cation resin and strong base anion resin/acrylic acid based weak base anion resin was also investigated. In the first case where strong base anion resin was used, there was precipitation in the system owing to the increase in pH while in the case where weak base anion resin was used there was no problem of precipitation and gadolinium removed effectively and the pH was around 6. (author)

  1. 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.

  2. Comparative study of plutonium burning in heavy and light water reactors

    International Nuclear Information System (INIS)

    There is interest in the U.S. and world-wide in reducing the burden on geological nuclear fuel disposal sites. In some disposal scenarios, the decay heat loading of the surrounding rock limits the commercial spent fuel capacity of the sites. In the long term (100 to 1,500 years), this decay heat is generated primarily by actinides, particularly 241Am and 241Pu. One possible approach to reducing this decay-heat burden would be to reprocess commercial spent nuclear fuel and use intermediate-tier thermal reactors to 'burn' these actinides and other transuranics (plutonium and higher actinides). The viability of this approach is dependent on the detailed changes in chemical and isotopic compositions of actinide-bearing fuels after irradiation in thermal reactor spectra. The intermediate-tier thermal burners could bridge the commercial water-cooled reactors and fast reactors required for ultimate consumption of the transuranics generated in the commercial reactors. This would reduce the number of such fast reactors required to complete the mission of burning transuranics. If thermal systems are to be used for the transmutation mission, it is likely that they would be similar to or are advanced versions of the systems currently used for power generation. In both the U.S. and Canada, light- and heavy-water-cooled thermal reactors are used for power generation in the commercial nuclear sector. About 103 pressurized- and boiling- light water reactors (PWRs and BRWs) are deployed in the U.S. nuclear industry while about 18 CANDU (heavy-water-cooled) reactors are used in the Canadian industry. There are substantial differences between light and heavy water-cooled reactors that might affect transmutation potential. These arise from differences in neutron balance of the reactors, in neutron energy spectra, in operational approaches (e.g., continuous refueling enhancing fuel burnup), and so on. A systematic study has been conducted to compare the transmutation potentials of

  3. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ionescu, S. [Institute for Nuclear Research Pitesti, Campului Str., 1, 115400 Mioveni (Romania)], E-mail: silviu.ionescu@nuclear.ro; Uta, O.; Parvan, M.; Ohai, D. [Institute for Nuclear Research Pitesti, Campului Str., 1, 115400 Mioveni (Romania)

    2009-03-31

    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.

  5. 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

  6. The key design features of the Indian advanced heavy water reactor

    International Nuclear Information System (INIS)

    The 235 MWe Indian Advanced Heavy Water Reactor (AHWR) is a vertical, pressure tube type, boiling light water cooled reactor. The three key specific features of design of the AHWR, having a large impact on its viability, safety and economics, relate to its reactor physics, coolant channel, and passive safety features. The reactor physics design is tuned for maximising use of thorium based fuel, and achieving a slightly negative void coefficient of reactivity. The fulfilment of these requirements has been possible through use of PuO2-ThO2 MOX, and ThO2 -U233O2 MOX in different pins of the same fuel cluster, and use of a heterogeneous moderator consisting of pyrolytic carbon and heavy water in 80%-20% volume ratio. The coolant channels of AHWR are designed for easy replaceability of pressure tubes, during normal maintenance shutdowns. The removal of pressure tube along with bottom end-fitting, using rolled joint detachment technology, can be done in AHWR coolant channels without disturbing the top end-fitting, tail pipe and feeder connections, and all other appendages of the coolant channel. The AHWR incorporates several passive safety features. These include core heat removal through natural circulation, direct injection of Emergency Core Coolant System (ECCS) water in fuel, passive systems for containment cooling and isolation, and availability of a large inventory of borated water in overhead Gravity Driven Water Pool (GDWP) to facilitate sustenance of core decay heat removal, ECCS injection, and containment cooling for three days without invoking any active systems or operator action. Incorporation of these features has been done together with considerable design simplifications, and elimination of several reactor grade equipment. A rigorous evaluation of feasibility of AHWR design concept has been completed. The economy enhancing aspects of its key design features are expected to compensate for relative complexity of the thorium fuel cycle activities required to

  7. Current developments and future challenges in physics analyses of the NRU heavy water research reactor

    International Nuclear Information System (INIS)

    The National Research Universal (NRU) reactor is heavy water cooled and moderated, with on-power fueling capability. TRIAD, a 3D two-group diffusion code, is currently used for support of day-to-day NRU operations. Recently, an MCNP full reactor model of NRU has been developed for benchmarking TRIAD. While reactivity changes and flux and power distributions from both methods are in reasonably good agreement, MCNP appears to eliminate a k-eff bias in TRIAD. Beyond TRIAD's capability, MCNP enables the assessment of radiation in the NRU outer structure. Challenges include improving TRIAD accuracy and MCNP performance, as well as performing NRU core-following using MCNP. (author)

  8. Some aspects of the thorium fuel cycle in heavy-water-moderated pressure tube reactors

    International Nuclear Information System (INIS)

    The use of thorium fuel cycles in heavy-water-moderated pressure tube (CANDU) reactors will allow much more energy to be extracted from a given amount of fuel than is possible with the present natural uranium cycle. The extent to which various factors affect thorium fuel cycle economics and resource consumption with equilibrium 233U levels in the fuel is considered. Resource consumption in growing nuclear power systems is also considered, and it is shown that considerable savings can be achieved even under conditions of rapid growth. The main elements of the development program necessary to provide the technological base for thorium fuel cycles in CANDU reactors are discussed. (author)

  9. Some aspects of the thorium fuel cycle in heavy-water-moderated pressure tube reactors

    International Nuclear Information System (INIS)

    The use of thorium fuel cycles in heavy-water-moderated pressure tube (CANDU) reactors will allow much more energy to be extracted from a given amount of fuel than is possible with the present natural uranium cycle. The extent to which various factors affect thorium fuel cycle economics and resource consumption with equilibrium 233U levels in the fuel is considered. Resource consumption in growing nuclear power systems is also considered, and it is shown that considerable savings can be achieved even under conditions of rapid growth. The main elements of the development program necessary to provide the technological base for thorium fuel cycles in CANDU reactors are discussed

  10. The controllability analysis of the purification system for heavy water reactors

    International Nuclear Information System (INIS)

    The heavy water reactor such as Wolsung No.1 and No.2 has a purification system to purify the reactor coolant. The control system regulates the coolant temperature to protect the ion exchanger. After the fuel exchanges of operating plant, the increase of the coolant pressure makes the purification temperature control difficult. In this paper, the controllability of the control dynamics of the purification system was analysed and the optimal parameters were proposed. To reduce the effects of the flow disturbance, the feedforward control structure was proposed and analysed

  11. 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

  12. 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.

  13. 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)

  14. 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

  15. 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

  16. Heavy-Water Reactors. Their Role in the Utilization of Long-Term Fuel Resources

    International Nuclear Information System (INIS)

    The paper discusses the merits of various composite reactor strategies, including several variants based entirely or partially on heavy-water reactors (HWRs). To meet the gradually increasing doubling time of the nuclear load predicted for the next half century the initial accent from the fuel conservation aspect should be on low-uranium inventory reactors and the ultimate accent on high-conversion ratio reactors. Optimum results are obtained by a gradual transition such as the following : Stage 1: Low-uranium inventory reactors only (e.g. HWR with nat. U feed and Pu recycle). Stage 2: Good Pu producer (e.g. HWR with nat. U) generating starting fissile inventory for reactor with higher conversion ratio and intermediate specific fissile inventory (e.g. HWR with thorium cycle optimized for conversion ratios somewhat below unity). Stage 3: As stage 2, but the latter reactor is optimized for still higher conversion (at expense of higher specific fissile inventory), e.g. thorium-cycle HWR optimized for conversion ratios of about unity. In the above programme, the role of the high-copversion reactor could also be taken by fast breeders as an alternative to the thorium cycle HWR. Both alternatives give comparable calculated uranium requirements for the projected demand developments for the next half century, and considerably lower uranium requirements than other combinations considered, e.g. light-water reactor/fast-breeder combinations or high-temperature gas-cooled reactor/fast-breeder combinations. The present worth of future integrated uranium costs for various strategies is also discussed. This provides strong financial incentives for the continued development of HWRs additional to the incentives arising from the improvement in HWR economics with the projected rise in unit sizes. Brief mention is made of the influence of the strategies on diffusion plant capacity requirements. (author)

  17. 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

  18. A study of the tritium behavior in coolant and moderator system of heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. P.; Song, S. S.; Chae, K. S. and others [Chosun Univ., Gwangju (Korea, Republic of)

    1993-12-15

    The objectives of this report is to present a regulatory policy on the environmental impact and personnel exposure by understanding the generation, accumulation, environmental release and management of tritium in heavy water reactors. By estimating the tritium concentration at Wolsong nuclear plant site by estimating and forecasting the generation and accumulation of tritium in coolant and moderator systems at Wolsong unit 1, we will study the management and release of tritium at Wolsong units 3 and 4 which are ready for construction. The major activities of this study are as follows : tritium generation and accumulation in heavy water reactor, a quantitative assessment of the accumulation and release of tritium at Wolsong nuclear plant site, heavy water management at Wolsong nuclear plants. The tritium concentration and accumulation trends in the systems at Wolsong unit 1 was estimated. A quantitative assessment of the tritium accumulation and release for Wolsong 2, 3 and 4 based on data from Wolsong 1 was performed. The tritium removal schemes and its long-term management plan were made.

  19. Sensitivity studies on nuclear data for thorium fuelled Advanced Heavy Water Reactor (AHWR)

    International Nuclear Information System (INIS)

    Sensitivity studies and uncertainty analyses on safety parameters for reactors are an important analysis tool for qualifying the basic nuclear cross-section data. It is also helpful in providing adequate margins at the design stage. In India, the design on Advanced Heavy water Reactor (AHWR) based on thorium is in its advanced stage of development. It is a first-of-a-kind reactor designed with many passive safety features which required to be qualified. In this paper, we discuss several types of sensitivity studies taken up for the integral parameters and reactivity coefficients for the AHWR-reference and the AHWR-LEU variant. The uncertainty studies are required to be taken to level higher where covariances can be established. It is important to analyse the uncertainties in a more rigorous manner

  20. 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

  1. 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.

  2. Photoneutron compensating method for boric acid concentration measuring instrument in heavy water moderated reactor

    International Nuclear Information System (INIS)

    In a boric acid concentration measuring instrument in a heavy water moderated reactor, a portion of γ-ray from Na-24 and Mn-56 is reacted with heavy water to form photoneutrons. The photoneutrons cause errors in the measurement for B-10 concentration. Then, in the present invention, a sample liquid containing photoneutron sources is supplied during normal measurement and a sample liquid removed with the photoneutron sources by passing through an ion exchange resin tower is supplied upon calibration of the measuring instrument. Then, the extent for the of effect of neutron sources and γ-nuclides is obtained by calculation from the measuring value to calibration the extent of the photoneutrons. Further, a method of using a counter tube having a Cd filter is used in combination during normal measurement to enable continuous measurement without exchanging the sample liquid. Accordingly, the influence of photoneutrons can be compensated and boric acid concentration can be measured at high accuracy. (N.H.)

  3. Research reactor core conversion from the use of highly enriched uranium to the use of low enriched uranium fuels. Guidebook addendum: Heavy water moderated reactors

    International Nuclear Information System (INIS)

    A Guidebook on Research Reactor Core Conversion from the Use of Highly Enriched Uranium to the Use of Low Enriched Uranium Fuels (IAEA-TECDOC--233) was issued by the International Atomic Energy Agency in August 1980. This document contains a wide variety of information of the physics, thermal-hydraulics, fuels, and fuel cycle economics for light water moderated research and test reactors. In consideration of the special features of heavy water moderated research and test reactors (hereafter referred to as heavy water research reactors), this Addendum to IAEA-TECDOC--233 has been prepared to assist operators and physicists from these reactors in determining whether conversion from HEU to LEU fuel designs is technically feasible for their specific reactor, and to assist in making a smooth transition to the use of LEU fuel designs where appropriate. The organization of this Addendum follows that of IAEA-TECDOC--233 as closely as possible in order to provide a consistent presentation of the information and to minimize the repetition of information that is common to both heavy water and light water research reactors. Distinctive features of the heavy water reactors are addressed where applicable

  4. 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)

  5. Burn up Analysis for Fuel Assembly Unit i n a Pressurized Heavy Water CANDU Reactor

    International Nuclear Information System (INIS)

    MCNPX code has been used for modeling a nd simulation of an assembly of CANDU Fuel bundle . The assembly is composed of a heterogeneous lattice of 37-element natural Uranium fuel, heavy water moderator and coolant. The fuel bundle is burnt in normal operation conditions of CANDU reactors. The effective multiplication factor (Keff ) of the bundle is calculated as a function of fuel burnup. The flux and power distribution are determined. Comparing t he concentrations of both Uranium and Plutonium isotopes are analyzed in the bundle. The results of the present model with the results of a benchmark problem, a good agreement was found PWR

  6. Transient subchannel analysis of turbine trip without bypass with IC of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Full text: The proposed advanced heavy water reactor (AHWR) is a 750 MWth vertical pressure tube type boiling light water cooled and heavy water moderated reactor. Safety analysis is carried out for various postulated initiating events (PIE) for AHWR. Turbine trip without bypass is one such PIE. Turbine trip results in rise in system pressure. Absence of bypass flow may lead to actuation of relief and safety valves depending upon system characteristics. System pressure rise activates various protective and engineered safeguard systems like reactor trip, isolation condenser and advanced accumulator limiting the consequences of the event. In this paper, a turbine trip without bypass is analysed, using a two-fluid code RELAPS/MOD3.2 [Fletcher, 1995] and subchannel analysis code COBRA IV-I. [Wheeler, 1976] Global analysis does not take into account the effect of different fuel pins with different peaking factors, effect of axial and radial cross flow mixing between adjacent subchannels. Combination of these two codes gives a better insight into the problem. The maximum rated reactor channel which houses 54 pin fuel bundle is modeled for COBRA IV-I simulation. The transient forcing function option of COBRA IV-I, validated by [Iwamura, 1994] with their flow-power transient experiments, has been used for transient thermal hydraulic parameter prediction for this study. Transient inlet pressure, inlet mass flow rate, inlet temperature, outlet enthalpy and the heat flux/power obtained from RELAPS /MOD3.2 simulation are among the boundary conditions employed on the COBRA-IV-I simulation. The thermal-hydraulic parameters predicted by RELAPS/MOD3.2 along with the effect of subchannel flow on the fuel element temperatures predicted by COBRA-IV-I are presented and discussed in this paper

  7. Heavy-water program: Technical requirements, conceptual design phase process, and reactor characteristics

    International Nuclear Information System (INIS)

    The Department of Energy (DOE), Office of New Production Reactors (NPRs), has initiated contracts with three architect-engineer/reactor vendors for the conceptual design phase (CDP) for new production reactors. This is the first major step of a procurement process that will proceed through final design, construction, testing, and operation of the facility. In preparation for the CDP for the NPRs, DOE developed a set of requirements that would serve as the basic technical reference throughout the entire life of the program-from the CDP to decommissioning. The objective was to establish requirements that, when met, will result in NPR capacity that will provide NPRs with design margins of sufficient magnitude to (1) provide the production capacity to supply goal quantities of tritium reliably and on a timely schedule; (2) provide a level of safety assurance that meets or exceeds that offered to the public by modern commercial nuclear power plants; (3) meet or exceed the applicable federal, state, and local regulations for safety and environmental requirements; (4) assure the availability of reactor, fuel, target, and processing technology for full-scale production on an urgent schedule with low schedule risk; and (5) meet the overall goal in the most cost-effective manner. The objectives of the CDP for the HWR [heavy-water reactor]-NPR are to provide comprehensively defined technical, cost, and schedule baselines for proceeding into preliminary design and to ensure that the technical and management competency of the CDP contractor team for conducting the preliminary design is demonstrated

  8. Analytical modelling and study of the stability characteristics of the Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    An analytical model has been developed to study the thermohydraulic and neutronic-coupled density-wave instability in the Indian Advanced Heavy Water Reactor (AHWR) which is a natural circulation pressure tube type boiling water reactor. The model considers a point kinetics model for the neutron dynamics and a lumped parameter model for the fuel thermal dynamics along with the conservation equations of mass, momentum and energy and equation of state for the coolant. In addition, to study the effect of neutron interactions between different parts of the core, the model considers a coupled multipoint kinetics equation in place of simple point kinetics equation. Linear stability theory was applied to reveal the instability of in-phase and out-of-phase modes in the boiling channels of the AHWR. The results indicate that the design configuration considered may experience both Ledinegg and Type I and Type II density-wave instabilities depending on the operating condition. Some methods of suppressing these instabilities were found out. In addition, it was found that the stability behavior of the reactor is greatly influenced by the void reactivity coefficient, fuel time constant, radial power distribution and channel inlet orificing. The delayed neutrons were found to have strong influence on the Type I and Type II instabilities. Decay ratio maps were predicted considering various operating parameters of the reactor, which are useful for its design. (author)

  9. Current developments and future challenges in physics analyses of the NRU heavy water research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, S.; Wilkin, B.; Leung, T., E-mail: nguyens@aecl.ca, E-mail: leungt@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2011-07-01

    The National Research Universal (NRU) reactor is heavy water cooled and moderated, with on-power fueling capability. TRIAD, a 3D two-group diffusion code, is currently used for support of day-to-day NRU operations. Recently, an MCNP full reactor model of NRU has been developed for benchmarking TRIAD. While reactivity changes and flux and power distributions from both methods are in reasonably good agreement, MCNP appears to eliminate a k-eff bias in TRIAD. Beyond TRIAD's capability, MCNP enables the assessment of radiation in the NRU outer structure. Challenges include improving TRIAD accuracy and MCNP performance, as well as performing NRU core-following using MCNP. (author)

  10. 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

  11. Criticality safety of fresh and irradiated fuel of the 6.5 MW heavy water research reactor

    International Nuclear Information System (INIS)

    Criticality safety problems, arising in away-from-reactor handling of the 6.5 MW heavy water research reactor fuel, are considered. The present status of fresh and spent fuel storage is described. To calculate criticality parameters of different non-reactor configurations of fresh and irradiated fuel, the well known WIMS code is combined with a few group three dimensional diffusion theory code TRITON. Typical results are presented and discussed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Jayalal, M.L., E-mail: jayalal@igcar.gov.in [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Ramachandran, Suja [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Rathakrishnan, S. [Reactor Physics Section, Madras Atomic Power Station (MAPS), Kalpakkam, Tamil Nadu (India); Satya Murty, S.A.V. [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Sai Baba, M. [Resources Management Group (RMG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India)

    2015-01-15

    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

  13. 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

  14. Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

    International Nuclear Information System (INIS)

    The irradiation of Th232 breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U238. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in the Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction β) for coolant voiding as standard NU fuel. (authors)

  15. 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

  16. Dosimetric Implications of Atmospheric Dispersal of Tritium Near a Heavy-water Research Reactor Facility

    International Nuclear Information System (INIS)

    An estimate of the tritium dose to the public in the vicinity of the heavy water research reactor facility at AECL-Chalk River Laboratories, Ontario, Canada, has largely been accomplished from analyses on regularly-collected samples of air, precipitation, drinking water and foodstuffs (pasture, fruit, vegetables and milk) and environmental dose models. To increase the confidence with which public doses are calculated, tritium doses were estimated directly from the ratio of tritiated species in urine samples from members of the general public. Single cumulative 24 h urine samples from a few adults living in the vicinity of the heavy-water research reactor facility at Chalk River Laboratories, Canada were collected and analysed for tritiated water and organically bound tritium. The participants were from Ottawa (200 km east), Deep River (10 km west) and Chalk River Laboratories. Tritiated water concentrations in urine ranged from 6.5 Bq.l-1 for the Ottawa resident to 15.9 Bq.l-1 for the Deep River resident, and were comparable to the ambient levels of tritium-in-precipitation at their locations. The ultra-low levels of organically bound tritium in urine from these same individuals were measured by 3He-ingrowth mass spectrometry and were 0.06 Bq.l-1 (Ottawa) and 0.29 Bq.l-1 (Deep River). For Chalk River Laboratories workers, tritiated water concentrations in urine ranged from 32 Bq.l-1 to 9.2x104 Bq.l-1, depending on the ambient levels of tritium in their workplace. The organically bound tritium concentrations in urine from the same workers were between 0.08 Bq.l-1 and 350 Bq.l-1. With a model based on the ratio of tritiated water to organically bound tritium in urine, the estimated dose arising from organically bound tritium in the body for the Ottawa and Deep River residents was about 26% and 50%, respectively, of the body water tritium dose. The workers in a reactor building at Chalk River Laboratories has less than 10% dose contribution from organically bound

  17. 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

  18. Safety Assessment of Pressure-Tube Heavy-Water Reactors by Probability Methods

    International Nuclear Information System (INIS)

    The use of probability methods for the assessment of safety has been developed in the United Kingdom for gas-cooled reactors; the development of these methods has proceeded in parallel with a move towards the formulation of quantitative safety criteria; some possible criteria are described. In particular, a method has been developed which is of value for a rapid assessment of a preliminary design in order to reveal potential points of weakness before the design is finalized. The method is of general application and in this paper it is applied to the design of hypothetical 250-MW(e) reactors of both the indirect and direct cycle types in order to illustrate the use of this method for pressure-tube heavy-water reactors and also to provide a comparison of the safety of this reactor system and others, such as the United Kingdom advanced gas-cooled type. A difficulty in the use of probability methods at present is the scarcity of data on failure rates for structures and large items of plant. An essential feature of the method described is a perturbation of the assumed failure rates by factors of order 100, to assess the effect of such uncertainties. The effect of major changes in other features important to safety, such as reliability of the containment systems, is also examined. (author)

  19. Thermo-mechanical behaviour of coolant channels for heavy water reactors under accident conditions

    International Nuclear Information System (INIS)

    The objective of nuclear safety research programme is to develop and verify computer models to accurately predict the behavior of reactor structural components under operating and off normal conditions. Indian Pressurised Heavy Water Reactors (PHWRs) are tube type of reactors. The coolant channel assemblies, being one of the most important components, need detailed analysis under all operating conditions as well as during postulated conditions of accidents for its thermo-mechanical behaviour. One of the postulated accident scenarios for heavy water moderated pressure tube type of reactors i.e. PHWRs is Loss Of Coolant Accident (LOCA) coincident with Loss Of Emergency Core Cooling System (LOECCS). In this case, even though the reactor is tripped, the decay heat may not be removed adequately due to low or no flow condition and inventory depletion of primary side. Since the emergency core cooling system is presumed to be not available, the cooling of the fuel pins and the coolant channel assembly depends on the moderator cooling system, which is assumed to be available. Moderator cooling system is a separate system in PHWRs. In PHWRs, the fuel assembly is surrounded by pressure tube, an annulus insulating environment and a concentric calandria tube. In this postulated accident scenario, a structural integrity evaluation has been carried out to assess the modes of deformation of pressure tube-calandria tube assembly in a tube type nuclear reactor. The loading of pressure and temperature causes the pressure tube to sag/balloon and come in contact with the outer cooler calandria tube. The resulting heat transfer could cool and thus control the deformation of the pressure tube thus introducing inter-dependency between thermal and mechanical contact behaviour. The amount of heat thus expelled significantly depends on the thermal contact conductance and the nature of contact between the two tubes. Deformation of pressure tube creates a heat removal path to the relatively

  20. Analysis of heavy water lattice experiments on research reactors for testing nuclear data

    International Nuclear Information System (INIS)

    There is a need for updated multigroup libraries for lattices codes of WIMS type for PHWR reactors calculations. Different multigroup libraries are used with WIMS and other codes, but these libraries are not normally updated to the level of last revision of ENDF/B-VI and other evaluated nuclear data files. Then, a special attention to the application of new WIMS libraries on PHWR calculations is justified. Some research and development activities associated to PHWR type of reactors, that need updated nuclear libraries of WIMS type, are: use of slightly enriched uranium (SEU cycle), use of UO2-ThO2 fuels, use of burnable poisons mixed in fuel pellets (UO2-Gd2O3) and absorber rods, new types of fuel elements (in Argentina: CARA Project-Advanced Fuel for Argentine Reactors) Taking into account the need of new WIMS libraries associated to these activities, a set of benchmarks have been identified and coded for PHWR lattice calculations.. The experimental benchmarks are identified with the name of the facility or research reactor where the measurements were carried out. The main references for this type of benchmarks is the ZED-2 Canadian reactor and DCA Japanese reactor. This work cover benchmark results of the following cases: ZED-2 analysis: experiments with 37 and 28 CANDU-type rod Fuel Clusters and lattice experiments with 19-rod Clusters with ThO2-UO2 Fuel; DCA analysis: Evaluation of Neutronic Parameters in Heavy Water and Slightly Enriched Uranium UO2 Fuel (28-rod Cluster) and critical experiments on Gadolinium poisoned cluster-type fuel assemblies of 54 rods in heavy water lattices of DCA facility. For several cases, results are included for different pitches and coolants. The parameters analysed are: k-effective with experimental bucklings, fast fission ratio [U-238 fissions/U-235 fissions], relative conversion ratio [U-238 captures/U-235 fissions], U-235 fission rate distribution, Cu-63 absorption rate distribution, Lutetium-Manganese activity ratio, ratio of

  1. 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

  2. Study of new structures adapted to gas-graphite and gas-heavy water reactors

    International Nuclear Information System (INIS)

    The experience acquired as a result of the operation of the Marcoule reactors and of the construction and start-up of the E.D.F. reactors on the one hand, and the conclusions of research and tests carried out out-of-pile on the other hand, lead to a considerable change in the general design of reactors of the gas-graphite type. The main modifications envisaged are analysed in the paper. The adoption of an annular fuel element and of a down-current cooling will make it possible to increase considerably the specific power and the power output of each channel; as a result there will be a considerable reduction in the number of the channels and a corresponding increase in the size of the unit cell. The graphite stack will have to be adapted to there new conditions. For security reasons, the use of prestressed concrete for the construction of the reactor vessel is becoming more widespread; they could lead to the exchangers and the fuel-handling apparatus becoming integrated inside the vessel (the so-called 'attic' device). A full-size mode) of this attic has been built at Saclay with the participation of EURATOM; the operational results obtained are presented as well as a new original design for the control rods. As for as the gas-heavy-water system is concerned, the research is carried out on two points of design; the first, which retains the use of horizontal pressure tubes, takes into account the experience acquired during the construction of the EL 4 reactor of which it will constitute an extrapolation; the second, arising from the research carried out on the gas-graphite system, will use a pre-stressed concrete vessel for holding the pressure, the moderator being almost at the same pressure as the cooling fluid and the fuel being placed in vertical channels. The relative merits of these two variants are analysed in the present paper. (authors)

  3. New techniques for the measurement of tritium activity releases in the air in heavy water reactors

    International Nuclear Information System (INIS)

    Tritium constitutes the principal potential source of ionizing radiation in heavy water reactors, playing the role of the dominant radionuclide of concern. In such reactors, the tritium hazard under normal working conditions is as important as all other radionuclide hazards combined. This calls for the employment of tritium monitoring systems capable of giving information about any releases in the shortest possible time so that immediate corrective measures could be adopted and operating personnel informed before they receive doses greater than permissible. Monitoring of tritium becomes complicated because of very low energy of tritium betas and presence of large concentrations of interfering activity and high background gamma fields inside the reactors. At the Rajasthan Atomic Power Station, for example, gamma compensated ionization chamber type of tritium monitors have been used but they have not been found entirely satisfactory. In this way, a period at least 30 minutes elapses before the level of tritium release could be known. This time delay in assessing the containment levels can prove critical in the event of acute releases and can result in excessive exposure to the operating personnel from tritium contamination. The need to undertake the present work arose. 4

  4. Definition and analysis of heavy water reactor benchmarks for testing new multigroup libraries

    International Nuclear Information System (INIS)

    A set of heavy water reactor benchmarks has been selected for testing new WIMS-D libraries. The libraries were constricted using data from ENDF/B-VI, Release 7, JENDL-3.2 and JEF-2.2 evaluated nuclear data files. The benchmarks cover a wide variety of reactor types and conditions, from fresh fuel to high burnup, and for natural and enriched uranium and Th-U fuels. The main parameters compared are the effective multiplication factor and other integral parameters, and isotopic composition of actinides on burnup cases. Besides, further investigations related with energy spectra used for preparation of WIMS-D libraries when applied on HWTR reactor calculations are included. Mostly of the benchmarks show a good agreement between experimental measurements and calculated values for all libraries. One exception is Th232 benchmark, were it is found that a library with JEND-3.2 Th232 data produces better results than ENDF/B-VI, R.7 and JEF-2.2 Th232 data. Results are slightly improved when HWTR spectra are used for weighting function to prepare the multi-group cross sections. This work is part of the International Atomic Energy Agency's Coordinated Research Project on 'Final Stage of WIMS-D Library Update Project'. (author)

  5. 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

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

    International Nuclear Information System (INIS)

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

  7. Core loading pattern optimization of thorium fueled heavy water breeder reactor using genetic algorithm

    International Nuclear Information System (INIS)

    In this work genetic algorithm was proposed to solve fuel loading pattern optimization problem in thorium fueled heavy water reactor. The objective function of optimization was to maximize the conversion ratio and minimize power peaking factor. Those objectives were simultaneously optimized using non-dominated Pareto-based population ranking optimal method. Members of non-dominated population were assigned selection probabilities based on their rankings in a manner similar to Baker's single criterion ranking selection procedure. A selected non-dominated member was bred through simple mutation or one-point crossover process to produce a new member. The genetic algorithm program was developed in FORTRAN 90 while neutronic calculation and analysis was done by COREBN code, a module of core burn-up calculation for SRAC. (authors)

  8. 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)

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

    International Nuclear Information System (INIS)

    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

  10. 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.

  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. Hydrodynamically induced dryout and post dryout important to heavy water reactors: A yearly progress report

    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 to study the CBF induced by various flow instabilities. Details of this test loop are presented. Initial test results have been presented. The two-phase flow regimes and hydrodynamic behaviors in the post dryout region have been studied under propagating rewetting conditions. The effect of subcooling and inlet velocity on flow transition as well as on the quench front propagation was investigated. The test liquid was Freon 113 which was introduced into the bottom of the quartz test section whose walls were maintained well above the film boiling temperature of the test liquid, via a transparent heat transfer fluid. The flow regimes observed down stream of the upward moving quench front were the rough wavy, the agitated, and the dispersed droplet/ligaments. A correlation for the flow regime transition between the inverted annular and the dispersed droplet/ligament flow patterns was developed. The correlation showed a marked dependence on the void fraction at the CBF location and hence on the flow regime encountered in the pre-CBF region

  13. High-conversion and high-burnup core concepts for pressure-tube-type heavy water reactors

    International Nuclear Information System (INIS)

    A high-conversion and a high-burnup core concept for a pressure-tube-type heavy water reactor are presented and analyzed from the standpoint of neutronics. These core concepts are based on the fact that neutron spectrum can be shifted by adjusting the amount of heavy water moderator outside the pressure tubes without affecting core-cooling capability. For the high-conversion core, where the heavy water moderator is replaced by a gas such as CO2 [carbon dioxide], a conversion ratio of more than 0.8 and an average discharge fuel burnup of 50GWd/t have been estimated to be attained with standard design fuel assemblies having 7.5% fissile Pu enrichment. For the high-burnup core, where fuel assemblies burned in the high-conversion (gas) region are relocated into the burner (heavy water) region, an average discharge fuel burnup of 110GWd/t has been estimated

  14. Daily tritium intakes by people living near a heavy-water research reactor facility: dosimetric significance

    International Nuclear Information System (INIS)

    We have estimated the relative daily intakes of tritiated water (HTO) and organically bound tritium (OBT), and have measured HTO-in-urine, in an adult population residing in the town of Deep River, Ontario, near a heavy-water research reactor facility at Chalk River. The daily intake of elevated levels of atmospheric tritium has been estimated from its concentration in environmental and biological samples, and various food items from a local tritium-monitoring program. Where the available data were inadequate, we used estimates generated by an environmental tritium-transfer model. From these data and estimates, we calculated a total daily tritium intake of about 55 Bq. Of this amount, 2.5 Bq is obtained from OBT-in-diet. Inhalation of HTO-in-air (15 Bq d-1) and HTO-in-drinking water (15 Bq d-1) accounts for more than half of the HTO intake. Skin absorption of HTO from air and bathing or swimming (for 30 min d-1) accounts for another 9 Bq d-1 and 0.1 Bq d-1, respectively. The remaining intake of HTO is from food as tissue-free water tritium. The International Commission on Radiological Protection's recommended two-compartment metabolic model for tritium predicts an equilibrium body burden of about 900 Bq from HTO (818 Bq) and OBT (83 Bq) in the body, which corresponds to an annual tritium dose of 0.41 μSv. The model-predicted urinary excretion of HTO (∼18 Bq L-1) agrees well with measured HTO-in-urine (range, 10-32 Bq L-1). The OBT dose contribution to the total tritium dose is about 16%. We conclude that for the people living near the Chalk River research reactor facility, the bulk of the tritium dose is due to HTO intake. (author)

  15. 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

  16. Inhalation radiotoxicity of irradiated thorium as a heavy water reactor fuel

    International Nuclear Information System (INIS)

    The online refueling capability of Heavy Water Reactors (HWRs), and their good neutron economy, allows a relatively high amount of neutron absorption in breeding materials to occur during normal fuel irradiation. This characteristic makes HWRs uniquely suited to the extraction of energy from thorium. In Canada, the toxicity and radiological protection methods dealing with personnel exposure to natural uranium (NU) spent fuel (SF) are well-established, but the corresponding methods for irradiated thorium fuel are not well known. This study uses software to compare the activity and toxicity of irradiated thorium fuel ('thorium SF') against those of NU. Thorium elements, contained in the inner eight elements of a heterogeneous high-burnup bundle having LEU (Low-enriched uranium) in the outer 35 elements, achieve a similar burnup to NU SF during its residence in a reactor, and the radiotoxicity due to fission products was found to be similar. However, due to the creation of such inhalation hazards as U-232 and Th-228, the radiotoxicity of thorium SF was almost double that of NU SF after sufficient time has passed for the decay of shorter-lived fission products. Current radio-protection methods for NU SF exposure are likely inadequate to estimate the internal dose to personnel to thorium SF, and an analysis of thorium in fecal samples is recommended to assess the internal dose from exposure to this fuel. (authors)

  17. Estimation of radiological source term from fuel following postulated LOCA in Indian pressurized heavy water reactors

    International Nuclear Information System (INIS)

    The estimation of source term from nuclear fuel is important for various radiological impact assessments including equipment qualification. This paper presents estimation of source term for Indian pressurized heavy water reactors (PHWRs) following postulated LOCA. A methodology has been developed for assessment of source term which considers effect of plant normal operating conditions, isotopic properties, distribution of the equilibrium core inventory due to power profile and geometric distribution of fuel mass and equilibrium core conditions. The source term takes into account the gap inventory on sheath failure and transient diffusional release predicted by in-house developed computer code 'STERCOR'. The estimations are carried out for selected radio-nuclides which contribute significantly to health effects upon release due to their quantities in reactor core, half-lives and emission properties. It is predicted that at the most 7.6% and 1.1% of the half core inventory of volatiles and noble gases respectively is released from fuel following postulated LOCA in 700 MWe. (author)

  18. Thorium-Fuelled Heavy-Water-Moderated Organic-Cooled Reactors

    International Nuclear Information System (INIS)

    The HWOCR is a heavy-water-moderated, organic-cooled, pressure-tube reactor similar to the ORGEL reactor concept being developed by Euratom. The performance of thorium fuel cycles in the HWOCR was evaluated and several 1000 MW(e) conceptual designs were developed for a thorium-fuelled HWOCR during an 18-month period from March 1965 to September 1966. The work was conducted in parallel with a much larger design and development effort on the uranium-fuelled concept and was restricted primarily to thorium recycle and core design aspects. The thorium fuel-cycle studies were performed under the USAEC's programme to develop economical, heavy-water reactors with optimum fuel utilization characteristics. During the initial stages, several potentially desirable combinations of fuel and clad materials were considered concurrently with investigations on compatible fuel assemblies and core geometries. Three different fuel assemblies were selected for further study: Zircaloy clad, coextruded thorium metal nested cylinders; vibratory compacted, SAP clad, thorium oxide pin bundles; and pelletized thorium monocarbide pin bundles. In the evolution of near-optimum conceptual designs, several cores featuring each of the three fuel assemblies were developed. As the work progressed, the initial goal to achieve optimum fuel utilization gradually shifted more toward consideration of economic performance. Except for relatively minor variations due to changes in the core arrangement or operating conditions, the uranium-fuelled HWOCR plant design was used in all the thorium-fuelled concepts. Three reference conceptual designs utilizing the three different fuel elements were completed and are based on extensive parameter studies and recycle considerations; the economics and fuel utilization of each design were evaluated, and technical feasibility, development costs, and compatibility with a uranium-optimized HWOCR were considered. A development programme was evolved that would lead to

  19. 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

  20. Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

    International Nuclear Information System (INIS)

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

  1. Preliminary Assessment of Heavy-Water Thorium Reactors in the Brazilian Nuclear Programme

    International Nuclear Information System (INIS)

    Since December 1965, the Instituto de Pesquisas Radioativas has been studying for the Brazilian Nuclear Energy Commission the feasibility of a thorium reactor programme in Brazil; since June 1966, the programme has been developed in close co-operation with the French Atomic Energy Commission. A reference conceptual design of a heavy-water-cooled and -moderated thorium converter reactor has been developed. The main features of that concept are the use of a prestressed-concrete pressure vessel, integrated arrangement of the primary circuit and the possibility of on-load fuel management. Economic competitiveness could be the result of high compactness, low capital costs and low fuel consumption. The technology involved is not very sophisticated; intensive engineering development work must be done in areas like fuel charge machine, concrete vessel insulation, and proper design of heat exchangers, but it is the feeling of the Group that these problems could be solved without seriously compromising the economic feasibility of the concept. Preliminary studies were made on the alternative use of enriched uranium or plutonium as a feed for the programme; in the latter case, plutonium could be produced in natural uranium reactors of the same type. The general conditions favouring each of these approaches to the thorium cycle have been determined, in particular those related to the costs of the fissile materials in the world market and to the country's policy related to nuclear fuel imports. The results of the preliminary studies are very encouraging and could justify the beginning of a research and development programme leading to the construction of a prototype in the 1970's. (author)

  2. Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

    International Nuclear Information System (INIS)

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (about 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)

  3. 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.

  4. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    International Nuclear Information System (INIS)

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  5. Construction of the core of the 'heavy water-gas' reactor EL 4

    International Nuclear Information System (INIS)

    The core of this reactor consists of a vessel containing heavy water, through which pass a series of pressure tubes for circulation of the cooling gas under boat pressure. The basic specifications which greatly influenced the design of this construction relate to aspects of safety in operation (fuel loading from both faces of the reactor, replacement of the components on both faces), neutronic demands (minimum absorption of the components lattice parameter, diameter of the pressure tubes) and thermal considerations (output temperature 500 C). These specifications have led to a' horizontal arrangement of the pressure tubes and raised very difficult problems of clearance, which make it impossible (for the dimensions of EL 4) to resort to expansion bellows on the pressure tubes. The result is a semi-rigid vessel in which the pressure tubes contribute to a large extent the mechanical resistance of the system by acting as a brace, whence the high stresses on the joints and pressure tubes (and the choice of zirconium alloys). The construction components include the pressure tube, the joints, the thermal insulation and the liner tube. A brief account is given of the testing methods used and the performances of these various units is particular. The safety factors foreseen for the pressure tube, and the design and manufacture, taking account of tolerances of the thickened ends necessary for fitting the tubes in place and designing the joints. The joints connecting the pressure tubes to the reactor tank, which are only accessible through the inside of the channel prolonging the pressure tube. These joints must not be a weak part in the construction. Two types have been developed: a rolled joint where the ends of the pressure tube are directly flanged onto the tank, and a welded joint using zircaloy-stainless steel transition pieces added to the ends of the pressure tube. All these joints are made by remote control and are removable. Two solutions have been found to the

  6. Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

    Science.gov (United States)

    Edwards, Geoffrey W R; Priest, Nicholas D

    2014-11-01

    The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low

  7. 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)

  8. 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)

  9. 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)

  10. Turning the volume down on heavy metals using tuned diatomite. A review of diatomite and modified diatomite for the extraction of heavy metals from water

    Energy Technology Data Exchange (ETDEWEB)

    Danil de Namor, Angela F., E-mail: A.Danil-De-Namor@surrey.ac.uk [Instituto Nacional de Tecnologia Industrial, Parque Tecnologico Industrial Miguelete, Buenos Aires (Argentina); Department of Chemistry, University of Surrey, Guildford, Surrey GU2 7XH (United Kingdom); El Gamouz, Abdelaziz [Department of Chemistry, University of Surrey, Guildford, Surrey GU2 7XH (United Kingdom); Frangie, Sofia; Martinez, Vanina; Valiente, Liliana [Instituto Nacional de Tecnologia Industrial, Parque Tecnologico Industrial Miguelete, Buenos Aires (Argentina); Webb, Oliver A. [Department of Chemistry, University of Surrey, Guildford, Surrey GU2 7XH (United Kingdom)

    2012-11-30

    Highlights: Black-Right-Pointing-Pointer Critical assessment of published work on raw and modified diatomites. Black-Right-Pointing-Pointer Counter-ion effect on the extraction of heavy metal speciation by diatomite. Black-Right-Pointing-Pointer Selection of the counter-ion by the use of existing thermodynamic data. Black-Right-Pointing-Pointer Enrichment of diatomites by attaching heavy metal selective functionalities. Black-Right-Pointing-Pointer Supramolecular chemistry for conferring selectivity to diatomites. - Abstract: Contamination of water by heavy metals is a global problem, to which an inexpensive and simple solution is required. Within this context the unique properties of diatomite and its abundance in many regions of the world have led to the current widespread interest in this material for water purification purposes. Defined sections on articles published on the use of raw and modified diatomite for the removal of heavy metal pollutants from water are critically reviewed. The capability of the materials as extracting agents for individual species and mixtures of heavy metals are considered in terms of the kinetics, the thermodynamics and the recyclability for both, the pollutant and the extracting material. The concept of 'selectivity' for the enrichment of naturally occurring materials such as diatomite through the introduction of suitable functionalities in their structure to target a given pollutant is emphasised. Suggestions for further research in this area are given.

  11. Remote dismantlement activities for the Argonne CP-5 Research Reactor

    International Nuclear Information System (INIS)

    The Department of Energy's (DOE's) Robotics Technology Development Program (RTDP) is participating in the dismantlement of a mothballed research reactor, Chicago Pile Number 5 (CP-5), at Argonne National Laboratory (ANL) to demonstrate technology developed by the program while assisting Argonne with their remote system needs. Equipment deployed for CP-5 activities includes the dual-arm work platform (DAWP), which will handle disassembly of reactor internals, and the RedZone Robotics-developed 'Rosie' remote work vehicle, which will perform size reduction of shield plugs, demolition of the biological shield, and waste packaging. Remote dismantlement tasks are scheduled to begin in February of 1997 and to continue through 1997 and beyond

  12. Development of advanced techniques for life management and inspection of advanced heavy water reactor (AWHR) coolant channel components

    International Nuclear Information System (INIS)

    Operating life of pressure tubes of Pressurized Heavy Water Reactor (PHWR) is limited due to the presence of various issues associated with the material like hydrogen pick up, delayed hydride cracking, axial elongation and increase in diameter due to irradiation creep and growth. Periodic monitoring of the health of the pressure tube under in-situ conditions is essential to ensure the safe operation of the reactor. New designs of reactor call for innovative design philosophy, modification in fabrication route of pressure tube, development of reactor specific tools, both analytical and hardware for assessing the fitness for service of the pressure tube. Feedback from existing reactors has enhanced the understanding about life limiting parameters. This paper gives an insight into the life limiting issues associated with pressure tube and the efforts pursued for development of life management techniques for coolant channel of Advanced Heavy Water Reactor (AHWR) designed in India. The tools and techniques for in-situ property/hydrogen measurement, pulsed eddy current technique for zirconium alloy in-homogeneity characterization, horizontal shear wave EMAT system for dissimilar metal weld inspection, sliver sampling of vertical channel etc. are elaborated in the paper. (author)

  13. Experimental thermal hydraulic facility for simulating LOCA behaviour of pressurised heavy water power reactor

    International Nuclear Information System (INIS)

    Experimental thermal hydraulic facility being set up adjacent to R and D Centre at Tarapur is a 13 MW full-elevation scaled down facility having the key components of PHT System of Pressurised Heavy Water Reactor (PHWR). The objective of the facility is to study thermal hydraulic behaviour of PHT System of PHWR by simulating various transients and accidental scenarios, to conduct safety related and operational transient studies and validation of various thermal hydraulic computer codes developed for analysis. The design of thermal hydraulic facility is based on the process parameters of a large PHWR with respect to fluid mass flux, transit time, flow velocity, pressure, temperature and enthalpy in PHT System. Experiments would be conducted in the facility to gain an improved understanding of the thermal hydraulic behaviour of large size PHWR during loss of coolant accident scenarios with forced and natural thermo-siphoning circulation modes etc. The data collected from the experiments would be used in validating computer codes developed for safety analysis. The facility is extensively instrumented to measure parameters such as temperature, pressure, flow, level, void-fraction at key locations. This paper gives the design philosophy used for scaling, design of major components of primary and secondary circuit of Experimental Thermal Hydraulic Facility and details of simulated experiments to be carried out. (author)

  14. Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

    International Nuclear Information System (INIS)

    Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior 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 to the commercial aircrafts operated in India

  15. 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.

  16. 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

  17. 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.

  18. General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel

    International Nuclear Information System (INIS)

    In the framework of research carried out on a CO2-cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors)

  19. 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

  20. Defense-in-depth evaluation model development strategy for pressurized heavy water reactor low power and shutdown operations

    International Nuclear Information System (INIS)

    The objective of defense-in-depth evaluation is to assess the level of defense-in-depth maintained during the various plant maintenance activities. Especially for shutdown and outage operations, the defense-in-depth might be challenged due to the reduction in redundancy and diversity resulting from the maintenance. Outage Risk Indicator of Nuclear Power Plants(ORION) which is a outage risk monitor for Korean NPPs is under development. ORION has the ability to assess the outage risk qualitatively using safety function assessment trees and deterministic margins such as time to boil. For pressurized heavy water reactors in Korea, defense-in-depth evaluation model development strategy to reflect the unique characteristics of pressurized heavy water reactors. The strategy and development steps were discussed in this paper

  1. 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...

  2. 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)

  3. 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

  4. 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

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

    International Nuclear Information System (INIS)

    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 design

  6. 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

  7. 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.

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

    International Nuclear Information System (INIS)

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

  9. Performance of refractometry in quantitative estimation of isotopic concentration of heavy water in nuclear reactor

    International Nuclear Information System (INIS)

    Highlights: ► Rapid analysis of heavy water samples, with precise temperature control. ► Entire composition range covered. ► Both variations in mole and wt.% of D2O in the heavy water sample studied. ► Standard error of calibration and prediction were estimated. - Abstract: The method of refractometry has been investigated for the quantitative estimation of isotopic concentration of heavy water (D2O) in a simulated water sample. Feasibility of refractometry as an excellent analytical technique for rapid and non-invasive determination of D2O concentration in water samples has been amply demonstrated. Temperature of the samples has been precisely controlled to eliminate the effect of temperature fluctuation on refractive index measurement. The method is found to exhibit a reasonable analytical response to its calibration performance over the purity range of 0–100% D2O. An accuracy of below ±1% in the measurement of isotopic purity of heavy water for the entire range could be achieved

  10. Fuel cycle flexibility in Advanced Heavy Water Reactor (AHWR) with the use of Th-LEU fuel

    International Nuclear Information System (INIS)

    The Advanced Heavy Water Reactor (AHWR) is being designed for large scale commercial utilization of thorium (Th) and integrated technological demonstration of the thorium cycle in India. The AHWR is a 920 MW(th), vertical pressure tube type cooled by boiling light water and moderated by heavy water. Heat removal through natural circulation and on-line fuelling are some of the salient features of AHWR design. The physics design of AHWR offers considerable flexibility to accommodate different kinds of fuel cycles. Our recent efforts have been directed towards a case study for the use of Th-LEU fuel cycle in a once-through mode. The discharged Uranium from Th-LEU cycle has proliferation resistant characteristics. This paper gives the initial core, fuel cycle characteristics and online refueling strategy of Th-LEU fuel in AHWR. (author)

  11. Initial operation of the Argonne superconducting heavy-ion linac

    International Nuclear Information System (INIS)

    Initial operation and recent development of the Argonne superconducting heavy-ion linac are discussed. The linac has been developed in order to demonstrate a cost-effective means of extending the performance of electrostatic tandem accelerators. The results of beam acceleration tests which began in June 1978 are described. At present 7 of a planned array of 22 resonators are operating on-line, and the linac system provides an effective accelerating potential of 7.5 MV. Although some technical problems remain, the level of performance and reliability is sufficient that appreciable beam time is becoming available to users

  12. 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)

  13. 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

  14. 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.

  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. 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)

  17. Analytical performance of refractometry in quantitative estimation of isotopic concentration of heavy water in nuclear reactor

    International Nuclear Information System (INIS)

    The method of refractometry has been investigated for the quantitative estimation of isotopic concentration of D2O (heavy water) in a simulated water sample. Viability of Refractometry as an excellent analytical technique for rapid and non-invasive determination of D2O concentration in water samples has been demonstrated. Temperature of the samples was precisely controlled to eliminate effect of temperature fluctuation on refractive index measurement. Calibration performance by this technique exhibited reasonable analytical response over a wide range (1-100%) of D2O concentration. (author)

  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. Application of the dose limitation system to the control of carbon-14 releases from heavy-water-moderated reactors

    International Nuclear Information System (INIS)

    Heavy-water-moderated reactors produce substantially more carbon-14 than light-water reactors. Applying the principles of the systems of dose limitation, the paper presents the rationale used for establishing the release limit for effluents containing this nuclide and for the decisions made regarding the effluent treatment in the third nuclear power station in Argentina. Production of carbon-14 in PHWR and the release routes are analysed in the light of the different effluent treatment possibilities. An optimization assessment is presented, taking into account effluent treatment and waste management costs, and the collective effective dose commitment due to the releases. The contribution of present carbon-14 releases to future individual doses is also analysed in the light of an upper bound for the contribution, representing a fraction of the individual dose limits. The paper presents the resulting requirements for the effluent treatment regarding carbon-14 and the corresponding regulatory aspects used in Argentina. (author)

  20. Preliminary definition of the design of a nuclear reactor for research and radioisotope production using natural uranium and heavy water

    International Nuclear Information System (INIS)

    A study was conducted about the evolution of the Brazilian importations of radioisotopes, from the beginning of the 70's since they have been increasingly used in the Country. In view of the limited production capacity of radioactive isotopes now existing in Brazil, a nuclear reactor type (natural uranium and heavy water) was defined, for research and production of radioisotopes, wich, besides providing, at least partially, the Brazilian needs of said isotopes, permits a large national participation in its project, construction and operating maintenance. The processes for heavy water production have been analyzed and it could be detected what is the best alternative for the production thereof, in low scale, in Brazil. The options concerning the definition of the main components of the reactor were justified and its most important features were determined, in relation to the neutronic and thermal aspects, being so defined its most significant parameters. The annual quantities were estimated, in terms of total and specific activity, for the radioisotopes that could be obtained by means of the proposed reactor, which, by now, are participating, to a large extent, in the total of Brazilian importation of radioactive isotopes. (Author)

  1. Validation of nuclear data for heavy water reactor lattices using WIMS and WIMKAL-88 nuclear data libraries

    International Nuclear Information System (INIS)

    Integral measurements of various types provide valuable data to assess the adequacy of the cross sections used in predicting the nuclear characteristics of reactors. In this context measurements of reactivity, relative reaction rates and neutron balance assume fundamental importance. We have analysed these parameters for heavy water moderated systems by using WIMS and WIMKAL-88 cross section libraries both of which have 69 energy groups. The analysis has been carried out by the lattice analysis code CLUB. It employs a method based on combination of interface current formalism and collision probability (CP) method. 6 refs, 1 tab

  2. 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.

  3. Modeling the effects of vapor pull through and entrainment in the simulation of stratified header in Pressurized Heavy water Reactors

    International Nuclear Information System (INIS)

    The Pressurized Heavy Water Reactors (PHWRs) form the mainstay of the Indian Nuclear Power Programme presently. Various components of PHWR like headers and channels have horizontal orientation. During accidents such as LOCA, flow through headers may get stratified for significant period of time under certain thermal hydraulic conditions like pressure, flow and quality. Determination of flow into reactor channels depend on the amount of vapour pull through/liquid entrainment from the stratified header. The paper describes an application of a model for simulation of stratified headers. The entrainment of liquid and pull through of steam in the feeder lines is calculated at different stratified flow levels of header at different pressures. A model has been developed based on Smoglies's model(1986) and used to predicted mass flow rate for branches located at different circumferential locations. The suitability of these relationships for adopting into integrated computer codes for accident analysis is discussed. (author)

  4. From a critical assembly heavy water - natural uranium to the fast - thermal research reactor in the Institute Vinca

    International Nuclear Information System (INIS)

    A part of the Institute in Vinca this monograph refers to is the thermal nuclear zero power reactor RB, with a heavy water moderator and variously enriched uranium fuel, that is, its present day version, the coupled fast-thermal system HERBE. A group of research workers, technicians, operators and skilled workmen in the workshop have worked continuously on it. Some of them have spent their whole working age at the reactor, and some a part of it. There is about a hundred and fifty internationally published papers, twenty master's and fourteen doctor's theses left behind them for the past thirty five years. This book is devoted to them. The first part of the text refers to the pioneering efforts on the reactor and fundamental research in reactor physics. The experimental reactor RB was designed and constructed at the time to operate with natural uranium and heavy water. Measurements are presented and the first results of reaching critical state, measurements of migration length of thermal neutrons and neutron multiplication factor in an infinite medium; also measurements of neutron flux density distribution and reactor parameter, and in the domain of safety, measurement of safety rods reactivity. Those were also the times when the known serious accident occurred with the uncontrolled rise of reactivity, which was especially minutely described in a publication of the International Atomic Energy Agency from Vienna. Later on, new fuel was acquired with 2 % enriched uranium. A series of experiments in reactor and neutron physics followed, with just the most interesting results of them presented here. In the period which followed, another type of fuel was available, with 80 % enriched uranium. New possibilities for work opened. Measurements with mixed lattices were performed, and the RA reactor lattices were simulated. After measurements mainly in the sphere of reactor and neutron physics, a need for investigations in the field of gamma and neutron radiation protection

  5. Numerical analysis and optimisation of heavy water upgrading column

    International Nuclear Information System (INIS)

    In the 'Pressurised Heavy Water' type of reactors, heavy water is used both as moderator and coolant. During operation of reactor downgraded heavy water is generated that needs to be upgraded for reuse in the reactor. When the isotopic purity of heavy water becomes less than 99.75%, it is termed as downgraded heavy water. Downgraded heavy water also contains impurity such as corrosion products, dirt, oil etc. Upgradation of downgraded heavy water is normally done in two steps: (i) Purification: In this step downgraded heavy water is first purified to remove corrosion products, dirt, oil, etc. and (ii) Upgradation of heavy water to increase its isotopic purity, this step is carried out by vacuum distillation of downgraded heavy water after purification. This project is aimed at mathematical modelling and numerical simulation of heavy water upgrading column. Modelling and simulation studies of the upgradation column are based on equilibrium stage model to evaluate the effect of feed location, pressure, feed composition, reflux ratio in the packed column for given reboiler and condenser duty of distillation column. State to stage modelling of two-phase two-component flow has constitutes the overall modelling of the column. The governing equations consist of stage-wise species and overall mass continuity and stage-wise energy balance. This results in tridigonal matrix equation for stage liquid fractions for heavy and light water. The stage-wise liquid flow rates and temperatures are governed by stage-wise mass and energy balance. The combined form of the corresponding governing equations, with the incorporation of thermodynamic equation of states, form a system of nonlinear equations. This system have been resolved numerically using modified Newton-Raphson method. A code in the MATLAB platform has been developed by on above numerical procedure. The optimisation of the column operating conditions is to be carried out based on parametric studies and analysis of different

  6. 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)

  7. Remote field eddy current technique for gap measurement of horizontal flux detector guide tube in pressurized heavy water reactor

    International Nuclear Information System (INIS)

    The fuel channels including the pressure tube(PT) and the calandria tube(CT) are important components of the pressurized heavy water reactor(PHWR). A sagging of fuel channel increases by heat and radiation exposure with the increasing operation time. The contact of fuel channel to the Horizontal flux Detector(HFD) guide tube is needed for the power plant safety. In order to solve this safety issue, the electromagnetic technique was applied to measure the status of the guide tube. The Horizontal flux Detector(HFD) guide tube and the Calandria tube(CT) in the Pressurized Heavy Water Reactor(PHWR) are cross-aligned horizontally. The remote field eddy current(RFEC) technology is applied for gap measurement between the HFD guide tube and the CT HFD guide tube can be detected by inserting the RFEC probe into pressure tube(PT) at the crossing point directly. The RFEC signals using the volume integral method(VIM) were simulated for obtaining the optimal inspection parameters. This paper shows that the simulated eddy current signals and the experimental results in variance with the CT/HFD gap.

  8. Recent development in ultra heavy section steel forgings for light water reactor components

    International Nuclear Information System (INIS)

    According to the tendency of reactor pressure vessels accompanying the increase of unit power output of light water reactors, the structural steel materials of very large size and thickness are demanded. In this paper, the performance, quality and manufacture of forgings with largest thickness are outlined. The features of reactor pressure vessels are the very large size of the vessels and their components, the very thick walls of shell flanges, and the manufacture of a shell flange and a nozzle belt in one body. Besides, the reduction of weld line and the adoption of assembling by circumferential welding, the reduction of the possibility of cracking in welds, and the improvements in low temperature toughness, high temperature strength and internal properties are required. As an example, the chemical composition, 400 ton ingot, heat treatment, mechanical properties, high temperature strength, and the quality of thick-walled portion of the shell flange for a 1300 MWe KWU type PWRPV are explained, and the NDT temperature is compared with that of the test plate for the HSST project in USA. The technology in this field in Japan is not behind that in foreign countries. (Kako, I.)

  9. An economic analysis of a light and heavy water moderated reactor synergy: burning americium using recycled uranium

    International Nuclear Information System (INIS)

    An economic analysis is presented for a proposed synergistic system between 2 nuclear utilities, one operating light water reactors (LWR) and another running a fleet of heavy water moderated reactors (HWR). Americium is partitioned from LWR spent nuclear fuel (SNF) to be transmuted in HWRs, with a consequent averted disposal cost to the LWR operator. In return, reprocessed uranium (RU) is supplied to the HWRs in sufficient quantities to support their operation both as power generators and americium burners. Two simplifying assumptions have been made. First, the economic value of RU is a linear function of the cost of fresh natural uranium (NU), and secondly, plutonium recycling for a third utility running a mixed oxide (MOX) fuelled reactor fleet has been already taking place, so that the extra cost of americium recycling is manageable. We conclude that, in order for this scenario to be economically attractive to the LWR operator, the averted disposal cost due to partitioning americium from LWR spent fuel must exceed 214 dollars per kg, comparable to estimates of the permanent disposal cost of the high level waste (HLW) from reprocessing spent LWR fuel. (authors)

  10. 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.

  11. 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.

  12. Defense-in-Depth risk evaluation model development strategy for plant configuration risk management of pressurized heavy water reactor low power/shutdown operation

    International Nuclear Information System (INIS)

    Configurations in nuclear power plants are defined by the outage status of plant equipment such as components, trains and systems. Equipment outage can occur by the maintenance or the unplanned equipment failures during power operation or low power/shutdown operation. The Configuration risk management plans were not developed for the low power/shutdown operation of pressurized heavy water reactors in Korea. In this study, the development strategy for the defense-in-depth risk evaluation model was developed as the part of the configuration risk management program for the low power/shutdown operation of pressurized heavy water reactors

  13. 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)

  14. Trend of R and D publications in pressurised heavy water reactors: A study using INIS and other databases

    International Nuclear Information System (INIS)

    Digital databases INIS (1970-2002), INSPEC (1969-2002), Chemical Abstracts (1977-2002), ISMEC (1973-June 2002), 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 Canadian researchers with researchers of other countries resulted in 75 publications. Among the most productive researchers in this field, the first 15 are from India. Top three contributors to PHWR publications with their respective authorship credits are: H.S. Kushwaha (106), Anil Kakodkar (100) and V. Venkat Raj (76). Prominent interdomainary interactions in PHWR subfields are: Specific nuclear reactors and associated plants with General studies of nuclear reactors (481), followed by Environmental sciences (185), and Materials science (154). Number of publications dealing with Geosciences aspect of environmental sciences are 141. Romania, Argentina, India and Republic of Korea have used mostly (≥75%) non-conventional media for publications. Out of the 4851 publications, 1228 have been published in 292 distinct journals. Top most journals publishing PHWR papers are: Radiation Protection and Environment (continued from: Bulletin of Radiation Protection since 1997), India (115); Nuclear Engineering International, UK (84); and Transactions of the American Nuclear Society, USA (68). (author)

  15. Derivation of Elastic Stress Concentration Factor Equations for Debris Fretting Flaws in Pressure Tubes of Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    If volumetric flaws such as bearing pad fretting flaws and debris fretting flaws are detected in the pressure tubes of pressurized heavy water reactors during in-service inspection, the initiation of fatigue cracks and delayed hydrogen cracking from the detected volumetric flaws shall be assessed by using elastic stress concentration factors in accordance with CSA N285.8-05. The CSA N285.8-05 presents only an approximate formula based on linear elastic fracture mechanics for the debris fretting flaw. In this study, an engineering formula considering the geometric characteristics of the debris fretting flaw in detail was derived using two-dimensional finite element analysis and Kinectrics, Inc.'s engineering procedure with slight modifications. Comparing the application results obtained using the derived formula with the three-dimensional finite element analysis results, it is found that the results obtained using the derived formula agree well with the results of the finite element analysis

  16. 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%.

  17. Review on the Management for Radioactive Effluent and Methodology for Setting of Derived Release Limits at Pressurized Heavy Water Reactors in Korea

    International Nuclear Information System (INIS)

    The radioactive effluents from pressurized heavy water reactors (PHWRs) are relatively larger than those from pressurized water reactors (PWRs). Futhermore, radioactive effluents from PHWRs are released continuously. Thus, the discharge of radioactive effluents is strictly controlled. To do this, radiation detectors are installed at stacks of reactor buildings to monitor the concentration of radioactive effluents in real-time. Derived release limits (DRLs) of annual discharge are also set up for each radionuclide and effluents are rigidly controlled not to exceed those limits. In this paper, the discharge process of radioactive effluents, the standard for establishment of DRL and its methodology, and currents status for PHWRs were reviewed

  18. Analysis of Metabolism and Effective Half-life for Tritium Intake of Radiation Workers at Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Tritium is the one of the dominant contributors to the internal radiation exposure of workers at pressurized heavy water reactors (PHWRs). This nuclide is likely to release to work places as tritiated water vapor (HTO) from a nuclear reactor and gets relatively easily into the body of workers by inhalation. Inhaled tritium usually reaches the equilibrium of concentration after approximately 2 hours inside the body and then is excreted from the body with a half-life of 10 days. Because tritium inside the body transports with body fluids, a whole body receives radiation exposure. Internal radiation exposure at PHWRs accounts for approximately 20-40% of total radiation exposure; most internal radiation exposure is attributed to tritium. Thus, tritium is an important nuclide to be necessarily monitored for the radiation management safety. In this paper, metabolism for tritium is established using its excretion rate results in urine samples of workers at PHWRs and an effective half-life, a key parameter to estimate the radiation exposure, was derived from these results. As a result, it was found that the effective half-life for workers at Korean nuclear power plants is shorter than that of International Commission on Radiological Protection guides, a half-life of 10 days

  19. 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.

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

    International Nuclear Information System (INIS)

    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

  1. Calorimeter measurements of absorbed doses at the heavy water enriched uranium reactor

    International Nuclear Information System (INIS)

    Application of calorimetry measurements of absorbed doses was imposed by the need of good knowledge of the absorbed dose values in the reactor experimental channels. Other methods are considered less reliable. The work was done in two phases: calorimetry measurements at lower reactor power (13-80 kW) by isothermal calorimeter, and differential calorimeter constructions for measurements at higher power levels (up to 1 MW). This report includes the following four annexes, papers: Isothermal calorimeter for reactor radiation monitoring, to be published; Calorimeter dosimetry of reactor radiation, presented at the Symposium about nuclear fuel held in april 1961; Radiation dosimetry of the reactor RA at Vinca, published in the Bull. Inst. Nucl. Sci. 1961; Differential calorimeter for reactor radiation dosimetry

  2. 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

  3. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes. Determination of hydrogen concentration and blister characterization

    International Nuclear Information System (INIS)

    Heavy water reactors (HWRs) comprise significant numbers of today's operating nuclear power plants, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems, especially pressure tubes, are an important factor in ensuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Project (CRP) on Intercomparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the framework of the IAEA'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 IAEA's project on advanced technologies for HWRs. The objective of the CRP was to 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 investigated the capability of different techniques to detect and characterize flaws. During the second phase participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in zirconium alloys. The intention was to identify the most effective pressure tube inspection and diagnostic methods and to identify further development needs. The organizations which participated in phase 2 of this CRP are: - Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL), Chalk River Laboratories (CRL), Canada; - Bhabha Atomic Research Centre (BARC), India; - Korea Atomic Energy Research Institute (KAERI), Republic of Korea; - National Institute for Research and Development for Technical Physics (NIRDTP), Romania; - Nuclear Non-Destructive Testing Research and Services (NNDT), Romania. IAEA-TECDOC-1499

  4. Dynamic analysis of containment building of 500 MWe pressurised heavy water reactor using finite element method

    International Nuclear Information System (INIS)

    The reactor building essentially comprises of a double containment system. The two containments viz. inner and outer containments (ICW and OCW), are axi-symmetric structures along with the raft and therefore an axi-symmetric finite element analysis of these should suffice. However, the inner containment is connected with the reactor internals (i.e. internal structures and calandria vault) at EL-130.0 meter elevation and thereby will have an interaction with the reactor internals. Exact modelling of this interaction effect is a formidable task since the reactor internals are not axi-symmetric. Hence, an equivalent axi-symmetric model of the reactor internals was evolved in such a way that the dynamic characteristics of the interaction effect are preserved. Analysis of the containment building has been carried out for two mutually perpendicular horizontal directions (N-S) and E-W) and the vertical direction using response spectrum and time history technique. Due credence was given to soil-structure interaction. This report presents the results and conclusions arrived at for these analyses. (author). 18 refs., 81 figs., 68 tabs., 3 appendixes

  5. 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. PMID:26081779

  6. Heavy water program: Technical requirements, conceptual design phase process and reactor characteristics

    International Nuclear Information System (INIS)

    In preparation for the Conceptual Design Phase (CDP) for the New Production Reactors (NPRs), the Department developed a set of requirements that would serve as the basic technical reference throughout the entire life of the program - from the CDP to decommissioning. The objective was to establish requirements which when met, will result in new production reactor capacity that will provide NPRs with design margins of sufficient magnitude: that will meet or exceed the level of safety assurance achieved by recent commercial reactors; that will provide a high degree of production assurance; that will achieve a level of environmental performance which will assure that requirements established in the NEPA process are met; and that will provide a high degree of confidence that the reactor will be constructed to meet urgent schedule needs. The NPR Requirements Document sets forth the management, technical, and implementing requirements for the CDP Process. Specifically, the implementing requirements define those contractor products (e.g., records, plans, analyses, studies, designs) to be developed and delivered during the CDP. The objectives of the CDP for the HWR-NPR are to provide comprehensively defined technical, cost and schedule baselines for proceeding into preliminary design and to ensure that the technical and management competency of the CDP contractor team for conducting the preliminary design is demonstrated. The mission to produce special nuclear materials by neutron induced transmutation of target materials has a major impact on the design of the reactor core and the reactor operating systems. This leads to the use of aluminum technology and, as a consequence, low temperature operation

  7. Verification Survey of the Building 315 Zero Power Reactor-6 Facility, Argonne National Laboratory-East, Argonne, Illinois

    Energy Technology Data Exchange (ETDEWEB)

    W. C. Adams

    2007-05-25

    Oak Ridge Institute for Science and Education (ORISE) conducted independent verification radiological survey activities at Argonne National Laboratory’s Building 315, Zero Power Reactor-6 facility in Argonne, Illinois. Independent verification survey activities included document and data reviews, alpha plus beta and gamma surface scans, alpha and beta surface activity measurements, and instrumentation comparisons. An interim letter report and a draft report, documenting the verification survey findings, were submitted to the DOE on November 8, 2006 and February 22, 2007, respectively (ORISE 2006b and 2007).

  8. Consideration of LH2 and LD2 cold neutron sources in heavy water reactor reflector

    International Nuclear Information System (INIS)

    The reactor power, the required CNS dimensions and power of the cryogenic equipment define the CNS type with maximized cold neutron production. Cold neutron fluxes from liquid hydrogen (LH2) and liquid deuterium (LD2) cold neutron sources (CNS) are analyzed. Different CNS volumes, presents and absence of reentrant holes inside the CNS, different adjustment of beam tube and containment are considered. (orig.)

  9. 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

  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. Heavy water reactor technology: The need for information exchange and increased international co-operation

    International Nuclear Information System (INIS)

    While PHWR technology has many favourable factors the reactor system is somewhat more complex with large number of pressure channels penetrating the reactor core. While many issues get resolved at the design stage itself, it is to be expected that some new issues always crop up at later stages. Expeditious resolution of such issues is important for the success of PHWR technology. It is rather unfortunate that information on PHWRs is rather limited in open literature as compared to the LWRs. We should therefore collectively work towards a regime where the technical information base is wider and such information can be exchanged freely to facilitate operation of PHWRs in an efficient and safe manner

  12. Experience of detecting blisters in irradiated coolant channels of Indian Pressurised Heavy Water Reactors

    International Nuclear Information System (INIS)

    A number of irradiated pressure tubes which were in contact with calandria tube during reactor operation have been subjected to detailed examination. In case of contact, calandria tube/ pressure tube (CT/PT) contact hydrogen absorbed in the pressure tube migrates and keeps accumulating in the contact region cold spot under thermal gradient. Over a length of time, accumulated hydrogen at the contact zone forms localized massive concentration of δ-phase zirconium hydride, which is termed as Blister. Blister grows in size with time in the reactor and reaches a critical size when it can crack. Presence of a cracked blister is a matter of concern for the safety of pressure tubes. Ultrasonic velocity ratio measurement technique has been developed and applied to evaluate formation of hydride blisters in irradiated pressure tube during the course of post irradiation examination. (author)

  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. Design, development and testing of 25 NB size Accumulator Isolation Passive Valve (AIPV) for Advanced Heavy Water Reactor (AHWR)

    International Nuclear Information System (INIS)

    In Advanced Heavy Water Reactor (AHWR), Emergency Core Cooling System (ECCS) is one of the engineered safety system provided to mitigate the consequences of Loss of Coolant Accident (LOCA) in the event of a break in the pressure boundary of Main Heat Transport (MHT) circuit. High Pressure Injection System of ECCS, is designed to provide coolant injection from advanced accumulators directly into the core for 15 minutes after LOCA. The injection pipe between each accumulator and ECCS header has a newly developed passive valve called Accumulator Isolation Passive Valve (AIPV). During normal reactor operation the MHT pressure will be 70 bar and accumulator pressure will be 55 bar. With rupture of large pipe, when the MHT system pressure falls down below 50 bar, the AIPV located between the accumulators and the ECC Headers, will open to provide coolant to the core. The AIPV is a self-acting type of valve requiring no external energy, i.e.neither air nor electric supply for its actuation. The AIPV serves not only as a passive isolation device but also as a flow control device. It is a non-standard, high pressure and high temperature valve and not manufactured by the valve industry worldwide. In the process of design and development of a 200 NB prototype AIPV for AHWR, a 25 NB size AIPV was designed and developed and successfully tested at Integral Test Loop (ITL). During several experiments carried out at ITL the functional capabilities of AIPV has been proved. The in-situ calibration and testing of AIPV in the plant without removing the same has also been established. This report deals with the role of AIPV in ECCS of AHWR, its design basis, tests performed at simulated conditions and test results with analysis. (author)

  15. 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

  16. 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

  17. Core design of heavy water cooled thorium breeder reactor with negative void reactivity and improved breeding performance

    International Nuclear Information System (INIS)

    A core of heavy water cooled thorium breeder reactor that produces 3.5 GWt energy using Th-233U oxide fuel has been studied to depict a concrete design specification. In order to improve the breeding performance compared to that of our previous study, one of key parameters in core design: moderator to fuel volume ratio (MFR) is re-surveyed. By reducing MFR from 1.0 to 0.6, the swing of keff during a cycle is considerably flattened, keeping negative void coefficient. The batch number is 3 and the refueling scheme employs out-in method to limit the radial power peaking factor less than 1.3. Due to efficient internal conversion, the reactivity of the core slightly increases with burnup, so that the cycle length is extended up to 1,300 days. Consequently, high averaged burnup of 80 GWd/t and breeding ratio of 1.07 at middle of cycle is achieved without any blankets. The number of control rods made of B4C is 19 and the total reactivity worth is -6.5% dk/k. The present core uses Zircaloy-4 as cladding material, the fast neutron fluence at EOC (End Of Cycle), however, exceeds its limit due to hard spectrum and long cycle length. As a part of future study, design will be further explored considering cladding integrity. (authors)

  18. 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)

  19. 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

  20. 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

  1. 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

  2. A Feasibility Study of Surveillance of Pressure Tubes in Heavy Water Reactors

    International Nuclear Information System (INIS)

    All the Zr-2.5Nb pressure tubes used in all CANDU 6 plants and Wolsong nuclear power plants were made according to AECL's design to have a strong tangential texture for high creep resistance. Nevertheless, they are to be replaced before reaching their design lifetime due to higher growth and diametral creep than expected, the latter of which causes a reduction of the plant power to below the 100% full power after 15 years of operation. These instances of operational experience show that AECL's design of pressure tubes is invalid. Besides, comparison of in-reactor creep between pressure tubes with radial and tangential textures was made not on the condition of all variables being constant except texture but with both texture and the microstructure being varied. In other words, the in-reactor creep tests carried out by AECL turns out to be ineffective to single out the effect of texture on creep. In contract, comparison of in-reactor creep between CANDU Zr-2.5Nb tube and Russian Zr-2.5Nb tubes demonstrate that creep of Zr-2.5Nb tubes are governed not by texture but by another factor. Given that Russian TMT-2 tube with the lower degree of tangential texture shows two times lower creep rate than CANDU Zr-2.5Nb tube, it is suggested that the stability of Nb dissolved in the α-Zr grains is a creep controlling factor. Fortunately, given our improved pressure tubes being developed in cooperation with Russia that were made similar to the TMT-2 tube's manufacturing process but optimized for a better control of fine precipitates and texture, it is evident that our improved tubes will be better due to excellent creep resistance, higher fracture toughness and zero axial growth when compared to TMT-2 tube. Since the 2005 version of the CSA N285.4 stipulates surveillance of pressure tubes, material examination of Wolsong Unit-2 pressure tubes should be conducted since 2013. Korea Hydro Nuclear Power (KHNP)'s strategy is to prepare an alternative instead of material examination

  3. Electrochemical removal of fluoride from water by PAOA-modified carbon felt electrodes in a continuous flow reactor.

    Science.gov (United States)

    Cui, Hao; Qian, Yan; An, Hao; Sun, Chencheng; Zhai, Jianping; Li, Qin

    2012-08-01

    A novel poly(aniline-co-o-aminophenol) (PAOA) modified carbon felt electrode reactor was designed and investigated for fluoride removal from aqueous solutions. This reactor design is innovative because it operates under a wider pH range because of coating with a copolymer PAOA ion exchange film. In addition, contaminant mass transfer from bulk solution to the electrode surface is enhanced by the porous carbon felt as an electron-conducting carrier material compared to other reactors. The electrically controlled anion exchange mechanism was investigated by X-ray photoelectron spectroscopy and cyclic voltammetry. The applicability of the reactor in the field was tested through a series of continuous flow experiments. When the flow rate and initial fluoride concentration were increased, the breakthrough curve became sharper, which lead to a decrease in the breakthrough time and the defluoridation capacity of the reactor. The terminal potential values largely influenced fluoride removal by the reactor and the optimal defluoridation efficiency was observed at around 1.2V. The breakthrough capacities were all >10mg/g over a wide pH range (pH 5-9) with an initial fluoride concentration of 10mg/L. Consecutive treatment-regeneration studies over a week (once each day) revealed that the PAOA-modified carbon felt electrode could be effectively regenerated for reuse. The PAOA-modified carbon felt electrode reactor is a promising system that could be made commercially available for fluoride removal from aqueous solutions in field applications. PMID:22595483

  4. 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

  5. Intercomparison and validation of computer codes for thermalhydraulic safety analysis of heavy water reactors

    International Nuclear Information System (INIS)

    Activities within the frame of the IAEA's Technical Working Group on Advanced Technologies for HWRs (TWG-HWR) are conducted in a project within the IAEA's subprogramme on nuclear power reactor technology development. The objective of the activities on HWRs is to foster, within the frame of the TWG-HWR, information exchange and co-operative research on technology development for current and future HWRs, with an emphasis on safety, economics and fuel resource sustainability. One of the activities recommended by the TWG-HWR was an international standard problem exercise entitled: Intercomparison and validation of computer codes for thermalhydraulics safety analyses. Intercomparison and validation of computer codes used in different countries for thermalhydraulics safety analyses will enhance the confidence in the predictions made by these codes. However, the intercomparison and validation exercise needs a set of reliable experimental data. The RD-14M Large-Loss Of Coolant Accident (LOCA) test B9401 simulating HWR LOCA behaviour that was conducted by Atomic Energy of Canada Ltd (AECL) was selected for this validation project. This report provides a comparison of the results obtained from six participating countries, utilizing four different computer codes. General conclusions are reached and recommendations made

  6. Comparison of Heavy Water Reactor Thermalhydraulic Code Predictions with Small Break LOCA Experimental Data

    International Nuclear Information System (INIS)

    Activities within the frame of the IAEA's Technical Working Group on Advanced Technologies for HWRs (TWG-HWR) are conducted in a project within the IAEA's subprogramme on nuclear power reactor technology development. The objective of the activities on HWRs is to foster, within the frame of the TWG-HWR, information exchange and cooperative research on technology development for current and future HWRs, with an emphasis on safety, economics and fuel resource sustainability. One of the activities recommended by the TWG-HWR was an international standard problem exercise entitled Intercomparison and Validation of Computer Codes for Thermalhydraulics Safety Analyses. Intercomparison and validation of computer codes used in different countries for thermalhydraulics safety analyses will enhance the confidence in the predictions made by these codes. However, the intercomparison and validation exercise needs a set of reliable experimental data. Two RD-14M small break loss of coolant accident (SBLOCA) tests, simulating HWR LOCA behaviour, conducted by Atomic Energy of Canada Ltd (AECL), were selected for this validation project. This report provides a comparison of the results obtained from eight participating organizations from six countries (Argentina, Canada, China, India, Republic of Korea, and Romania), utilizing four different computer codes (ATMIKA, CATHENA, MARS-KS, and RELAP5). General conclusions are reached and recommendations made.

  7. Structural Analysis of Surface-Modified Oxidation-Resistant Zirconium Alloy Cladding for Light Water Reactors

    International Nuclear Information System (INIS)

    While the current zirconium-based alloy cladding (Zircaloy, here after) has served well for fission-product barrier and heat transfer medium for the nuclear fuel of light water reactors (LWRs) in steady-states, concerns surrounding its mechanical behavior during accidents have drawn serious attentions. In accidents, strength degradation of the current-zirconium based alloy cladding manifests at temperature around ∼800 .deg. C, which results in fuel ballooning. Above 1000 .deg. C, zircaloy undergoes rapid oxidation with steam. Formation of brittle oxide (ZrO2) and underlying oxygen-saturated α-zircaloy as a consequence of steam oxidation leads to loss of cladding ductility. Indeed, the loss of zircaloy ductility upon the steam oxidation has been taken as a measure of fuel failure criteria as stated in 10 CFR 50.46. In addition, zircaloy steam oxidation is an exothermic reaction, which is an energy source that sharply accelerates temperature increase under loss of coolant accidents, decreasing allowable coping time for loss of coolant accidents, decreasing allowable coping time for loss of coolant accidents (LOCA) before significant fuel/core melting starts. Hydrogen generated as a result of zircaloy oxidation could cause an explosion if certain conditions are met. In steady-state operation, zircaloy embrittlement limits the burnup of the fuel rod to assure remaining cladding ductility to cope with accidents. As a way to suppress hydrogen generation and cladding embrittlement by oxidation, ideas of cladding coating with an oxidation-preventive layer have emerged. Indeed, among a numbers of 'accident-tolerant-fuel (ATF)' concepts, the concept of coating the current fuel rod. Some signs of success on the lab-scale oxidation-prevention have been confirmed with a few coating candidates. Yet, relatively less attention has been given to structural integrity of coated zirconium-based alloy cladding. It is important to note that oxidation-suppression performance

  8. Study on remodeling the heavy water facility of the Kyoto University reactor for neutron capture therapy from the concept of neutron energy spectrum control

    International Nuclear Information System (INIS)

    In August 1988, there was heavy water leakage from the thermocouple guide pipe of the heavy water tank adjacent to the core of the Kyoto University Research Reactor (KUR). The need for a fundamental reexamination of the Heavy Water Thermal Neutron Facility (HWTNF) has been recognized since the guide pipe was repaired. Clinical irradiation was restarted in February 1990, and as of September 1, 1992, 24 clinical treatments had been carried out. Some improvements for the clinical treatment have been requested by clinicians and other users, especially the ability to use the facility while the reactor is in continuous operations. From the viewpoints of reactor engineering and medical physics, there are four goals for remodeling the present HWTNF: (1) to simplify and secure maintenance, overhaul, and repair, (2) to enable the facility to be used during continuous operation of the KUR, (3) to improve the performances of the irradiation field of thermal neutrons for biomedical uses, and (4) to control neutron energy spectrum. In this paper, an outline for the redesign of the facility is reported

  9. 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

  10. 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)

  11. Some considerations on utilizing a Canadian heavy-water reactor in a dual-purpose power-desalination plant

    International Nuclear Information System (INIS)

    In the paper two approaches are illustrated: Use of Douglas Point design with a condensing turbine to produce mainly power but with some steam extracted to produce a moderate amount of desalted water; use of typical Pickering reactor with a back pressure turbine to produce a large quantity of desalinated water and a moderate amount of power. 2 figs, 6 tabs

  12. 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

  13. 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.

  14. 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.

  15. Role of Passive Safety Features in Prevention And Mitigation of Severe Plant Conditions in Indian Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation

  16. 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

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

    International Nuclear Information System (INIS)

    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

  18. 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.

  19. 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

  20. 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

  1. Fluid-phase transitions in light water and heavy water

    International Nuclear Information System (INIS)

    Phase transitions are important phenomena in water-cooled nuclear-fission reactors. In this study, liquid-vapour phase transitions and critical-point behaviour are considered. Special attention is given to the van der Waals model, asymptotic critical scaling, thermodynamic potentials and steam properties. Models and data are shown for light water and heavy water. The relevance to advanced Generation III reactors and proposed Generation IV reactors is assessed. (author)

  2. 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

  3. Trend of R and D publications in pressurised heavy water reactors: A study using INIS and other databases

    International Nuclear Information System (INIS)

    Digital databases INIS, INSPEC, ISMEC, Chemical Abstracts, Science Citation Index, Web of science, Chemistry Citation Index, BIOSIS, Medline and Analytical Abstracts were used for comprehensive retrieval of bibliographic details of publications on Pressurised Heavy Water Reactors (PHWR) research. Keyword search mechanism adopted for searches in title or descriptors resulted in 5863 records. Taking INIS as base the duplicate records within INIS and between INIS and other databases were identified and removed. Remaining 4851 records were considered for further study. A manual examination of Abstracts of randomly selected 500 records among 4851 showed that about 3% records were not directly related to PHWR research but only remotely related. It is assumed that this would not adversely affect the result of this study. A detailed analysis of 4851 records was carried out to examine country wise publications, contributing authors, interdomainary interactions, preferred media for publication etc. Out of the 4851 records, 196 distinct records of publications could not be found in INIS but in other scientific databases, mainly in INSPEC (117 records) and Chemical Abstract (63 records). Year-wise growth of R and D publications related to PHWR since year 1966 is depicted. The curve gives some indications of Ideal logistic growth model which states that logistic growth in any field of knowledge ideally takes an extended S-shape. Also continued escalation in growth after expected maturation implies emergence of new directions in research, new discoveries and the new opportunities. Among the 46 countries contributing to PHWR research, India with 1737 publications is the forerunner followed by Canada (1492), Romania (508) and Argentina (334). The total literature output on PHWR research from top 6 countries amounted to about 90%. A graphical representation of the history of ten top contributing countries is presented. With respect to the international collaboration Canada is at

  4. 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

  5. Indian heavy water programme - challenges and opportunities

    International Nuclear Information System (INIS)

    Discovery of fission of uranium in 1939 opened up hitherto unknown possibilities for utilising the fission energy for use of mankind, mainly for the production of and electrical energy. It was realised that this nuclear energy could be an ideal substitute for the fast depleting fossil fuels which would one day get exhausted. Two main concepts of nuclear power reactor got evolved, one enriched uranium fuelled, ordinary water moderated reactor and another natural uranium fuelled heavy water moderated reactor. The concentration of uranium 235U needed for ordinary water moderated reactors is 3% but the naturally occurring uranium in India contains only 0.7% of 235U. The reactors utilising natural uranium as fuel require Heavy Water as moderator. The processing of uranium ore to achieve from 0.7% to 3% is highly complex. Recognising the fact that India has limited uranium resources but rich thorium resources, Dr. Bhabha formulated a three stage nuclear power generation programme for our country. The first generation reactors can use natural uranium as fuel with heavy water as moderator. Since the technology to generate such large scale heavy water to match the urgent need for nuclear power generation was not indigenously available, the technology available with Canada and France was utilised for installation of first generation heavy water plants in India. However, the peaceful nuclear experiment conducted by India in 1974 caused resentment among the countries that supplied Heavy Water technology to India and they stopped all technological help and assistance in nuclear field. Thereafter, it was the story of India going alone in heavy water production. That made India meets successfully all challenges on the way to installation, commissioning and sustained operation of all plants. Today we have six operating Heavy Water plants, spread all over the country. We have reached a stage, a change from a situation of crunch to a level of not only self sufficiency but to a state

  6. Heavy water wastes purification from tritium by CECE process

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, I.A.; Bondarenko, S.D.; Fedorchenko, O.A.; Vasyanina, T.V.; Konoplev, K.A.; Arkhipov, E.A. [Petersburg Nuclear Physics Institute, Leningrad (Russian Federation); Uborsky, V.V. [JSC ' DOL' , Moscow (Russian Federation)

    2007-07-01

    Future fusion reactors require Isotope Separation System for tritium extracting mainly from light water. Nuclear reactors moderated by heavy water also require upgrading facility to maintain deuterium concentration in water and facility for tritium recovery. The problems of tritium removal from heavy and light water and upgrading of tritiated heavy water wastes are issue of the day as before. To date the combined electrolysis catalytic exchange (CECE) process utilizing wetproofed catalyst is the most attractive one for extracting tritium from water due to its high separation factors and near-ambient operating conditions. The experimental industrial plant has been built in PNPI for the development of the CECE technology for hydrogen isotope separation. The process uses a LPCE column and electrolysis cells to convert water to hydrogen. The plant has been in operation about 10 years. In parallel with a development of CECE process for hydrogen isotope separation the plant is used for reprocessing tritium heavy water waste. Processing waste with the content of {proportional_to} 47 % of heavy hydrogen and 108 Bq/kg of tritium, the plant produces 99.85-99.995% heavy water and deuterium gas for science and industry. Owing to industrial demands for heavy water with reduced tritium content, the plant was modified and additional equipment and procedures were put in place to operate in the detritiation mode. After prolonged operation campaigns it was decided to update the plant with an additional separation column connected with existing equipment. Now the main parts of plant are two 100-diameter exchange columns of 7.5 m and 6.9 m overall height correspondingly, alkaline electrolytic cells. The columns are filled with alternating layers of wetproofed catalyst developed by Mendeleev University and stainless steel spiral-prismatic packing. The first column consists of five separation sections connected through a distributor of liquid, the second column consists of three

  7. Heavy water wastes purification from tritium by CECE process

    International Nuclear Information System (INIS)

    Future fusion reactors require Isotope Separation System for tritium extracting mainly from light water. Nuclear reactors moderated by heavy water also require upgrading facility to maintain deuterium concentration in water and facility for tritium recovery. The problems of tritium removal from heavy and light water and upgrading of tritiated heavy water wastes are issue of the day as before. To date the combined electrolysis catalytic exchange (CECE) process utilizing wetproofed catalyst is the most attractive one for extracting tritium from water due to its high separation factors and near-ambient operating conditions. The experimental industrial plant has been built in PNPI for the development of the CECE technology for hydrogen isotope separation. The process uses a LPCE column and electrolysis cells to convert water to hydrogen. The plant has been in operation about 10 years. In parallel with a development of CECE process for hydrogen isotope separation the plant is used for reprocessing tritium heavy water waste. Processing waste with the content of ∝ 47 % of heavy hydrogen and 108 Bq/kg of tritium, the plant produces 99.85-99.995% heavy water and deuterium gas for science and industry. Owing to industrial demands for heavy water with reduced tritium content, the plant was modified and additional equipment and procedures were put in place to operate in the detritiation mode. After prolonged operation campaigns it was decided to update the plant with an additional separation column connected with existing equipment. Now the main parts of plant are two 100-diameter exchange columns of 7.5 m and 6.9 m overall height correspondingly, alkaline electrolytic cells. The columns are filled with alternating layers of wetproofed catalyst developed by Mendeleev University and stainless steel spiral-prismatic packing. The first column consists of five separation sections connected through a distributor of liquid, the second column consists of three separation sections

  8. 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

  9. Heavy water at Aswan

    International Nuclear Information System (INIS)

    A fertilizer factory is being built by Egyptian Chemical Industries (Kima) at Aswan on the upper Nile; it will produce a mixture of ammonium nitrate and calcium carbonate adjusted to contain 20.5% nitrogen. It is also proposed to construct a heavy water plant to be located at and integrated with the fertilizer factory. At the request of the Government of the United Arab Republic, the International Atomic Energy Agency sent an expert to carry out investigation of the technical, economic and other related aspects of the proposed production of heavy water. A report was submitted to the IAEA Director General. Its main conclusions can be summarized as follows: (1) Production of heavy water as a by-product of fertilizer manufacture at Aswan is technically feasible. Separation of deuterium from industrial hydrogen for this purpose could be done either by catalytic exchange or by liquefaction and distillation; the choice should depend on economic considerations. (2) The heavy water produced at Aswan should be competitive in cost with that produced elsewhere; this, however, would depend on whether firm contracts are obtained for the delivery of equipment at guaranteed prices and with guaranteed performance, and whether such prices are in reasonable agreement with preliminary estimates. (3) The future market for heavy water is difficult to predict. For one thing, there is a very large production capacity in the USA, most of which is idle due to lack of demand. Secondly, there is a relatively small production outside the USA that is sold at prices higher than that charged by the US Government. The future of the market is necessarily contingent upon the possibility of future free sale by the US Government. At the end of his report, the expert has also given his comments on possible further assistance to the project by IAEA

  10. Verification Survey of the Building 315 Zero Power Reactor-6 Facility, Argonne National Laboratory-East, Argonne, Illinois

    International Nuclear Information System (INIS)

    Oak Ridge Institute for Science and Education (ORISE) conducted independent verification radiological survey activities at Argonne National Laboratory's Building 315, Zero Power Reactor-6 facility in Argonne, Illinois. Independent verification survey activities included document and data reviews, alpha plus beta and gamma surface scans, alpha and beta surface activity measurements, and instrumentation comparisons. An interim letter report and a draft report, documenting the verification survey findings, were submitted to the DOE on November 8, 2006 and February 22, 2007, respectively (ORISE 2006b and 2007). Argonne National Laboratory-East (ANL-E) is owned by the U.S. Department of Energy (DOE) and is operated under a contract with the University of Chicago. Fundamental and applied research in the physical, biomedical, and environmental sciences are conducted at ANL-E and the laboratory serves as a major center of energy research and development. Building 315, which was completed in 1962, contained two cells, Cells 5 and 4, for holding Zero Power Reactor (ZPR)-6 and ZPR-9, respectively. These reactors were built to increase the knowledge and understanding of fast reactor technology. ZPR-6 was also referred to as the Fast Critical Facility and focused on fast reactor studies for civilian power production. ZPR-9 was used for nuclear rocket and fast reactor studies. In 1967, the reactors were converted for plutonium use. The reactors operated from the mid-1960's until 1982 when they were both shut down. Low levels of radioactivity were expected to be present due to the operating power levels of the ZPR's being restricted to well below 1,000 watts. To evaluate the presence of radiological contamination, DOE characterized the ZPRs in 2001. Currently, the Melt Attack and Coolability Experiments (MACE) and Melt Coolability and Concrete Interaction (MCCI) Experiments are being conducted in Cell 4 where the ZPR-9 is located (ANL 2002 and 2006). ANL has performed final

  11. Decontamination and decommissioning of the JANUS reactor at the Argonne National Laboratory-East site

    International Nuclear Information System (INIS)

    Argonne National Laboratory has begun the decontamination and decommissioning (D ampersand D) of the JANUS Reactor Facility. The project is managed by the Technology Development Division's D ampersand D Program personnel. D ampersand D procedures are performed by sub-contractor personnel. Specific activities involving the removal, size reduction, and packaging of radioactive components and facilities are discussed

  12. Heavy water at Aswan

    International Nuclear Information System (INIS)

    In response to a request from the Government of the United Arab Republic, two experts were sent by IAEA, at the beginning of November, to Egypt, Mr. B.V. Nevsky (USSR) and Dr. Thayer (USA). Mr. Nevsky examined the possibilities and economics of extracting uranium from ores rich in phosphates, which are plentiful in Egypt. His report was not available when the Bulletin went to press. Mr. Thayer has reported to the Director General on the possibilities and the economic interest of producing heavy water by the electrolytic method as a by-product of an ammonium nitrate fertilizer factory now being constructed at Aswan. Dr. Thayer is a member of the Atomic Energy Division of the Du Pont Company and is working as a heavy water expert at the American Atomic Energy Commission's Savannah River project. We publish below excerpts from Dr. Thayer's report

  13. Liquid sloshing in gravity driven water pool of Advanced Heavy Water Reactor - pool liquid under design seismic load and slosh control studies

    International Nuclear Information System (INIS)

    Sloshing phenomenon is well understood in regular cylindrical and rectangular liquid tanks subjected to earthquake. However, seismic behaviour of water in complex geometry such as a sectored annular tank, e.g., Gravity Driven Water Pool (GDWP) which is located in Advanced Heavy Water Reactor (AHWR) need to be investigated in detail in the view of safety significance. Initially, for validation of Computational Fluid Dynamics (CFD) procedure, square and four sectored square tanks are taken. Slosh height and liquid pressure are calculated over time through theoretical and experimental procedures. Results from theoretical and experimental approaches are compared with CFD results and found to be in agreement. The present work has two main objectives. The first one is to investigate the sloshing behaviour in an un-baffled and baffled three dimensional single sector of GDWP of AHWR under sinusoidal excitation. Other one is to study the sloshing in GDWP water using simulated seismic load along the three orthogonal directions. This simulated seismic load is generated from design basis floor response spectrum data (FRS) of AHWR building. For this, the annular tank is modelled along with water and numerical simulation is carried out. The sinusoidal and earthquake excitations are applied as acceleration force along with gravity. For the earthquake case, acceleration-time history is generated compatible to the design FRS of AHWR building. The free surface is captured by Volume of Fluid (VOF) technique and the fluid domain is solved by finite volume method while the structural domain is solved by finite element approach. Un-baffled and baffled tank configurations are compared to show the reduction in wave height under excitation. The interaction between the fluid and pool wall deformation is simulated using a partitioned fluid-structure coupling. In the earthquake case, a user subroutine function is developed to convert FRS in to time history of acceleration in three directions

  14. 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.)

  15. 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)

  16. 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.

  17. Light water type reactor

    International Nuclear Information System (INIS)

    The nuclear reactor of the present invention prevents disruption of a reactor core even in a case of occurrence of entire AC power loss event, and even if a reactor core disruption should occur, it prevents a rupture of the reactor container due to excess heating. That is, a high pressure water injection system and a low pressure water injection system operated by a diesel engine are disposed in the reactor building in addition to an emergency core cooling system. With such a constitution, even if an entire AC power loss event should occur, water can surely be injected to the reactor thereby enabling to prevent the rupture of the reactor core. Even if it should be ruptured, water can be sprayed to the reactor container by the low pressure water injection system. Further, if each of water injection pumps of the high pressure water injection system and the low pressure water injection system can be driven also by motors in addition to the diesel engine, the pump operation can be conducted more certainly and integrally. (I.S.)

  18. 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

  19. 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.

  20. 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

  1. 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

  2. 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)

  3. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    The description is given of a water cooled nuclear reactor comprising a core, cooling water that rises through the core, vertical guide tubes located inside the core and control rods vertically mobile in the guide tubes. In this reactor the cooling water is divided into a first part introduced at the bottom end of the core and rising through it and a second part introduced at the top end of the guide tubes so as to drop in them

  4. 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

  5. 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.

  6. The Bare Critical Assembly of Natural Uranium and Heavy Water

    International Nuclear Information System (INIS)

    The first reactor built in Yugoslavia was the bare zero energy heavy water and natural uranium assembly at the Boris Kidric Institute of Nuclear Sciences, Belgrade. The reactor went critical on April 29, 1958. The possession of four tons of natural uranium metal and the temporary availability of seven tons of heavy water encouraged the staff of the Institute to build a critical assembly. A critical assembly was chosen, rather than high flux reactor, because the heavy water was available only temporarily. Besides, a 10 MW, enriched uranium, research reactor is being built at the same Institute and should be ready for operation late this year. It was supposed that the zero energy reactor would provide experience in carrying out critical experiments, operational experience with nuclear reactors, and the possibility for an extensive program in reactor physics. (author)

  7. Reactor water sampling device

    International Nuclear Information System (INIS)

    The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

  8. Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors (U)

    International Nuclear Information System (INIS)

    This paper reports on a full-scope probabilistic risk assessment (PRA) performed for the Savannah River Site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident

  9. 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

  10. Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor

    International Nuclear Information System (INIS)

    Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors)

  11. 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

  12. 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

  13. 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

  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. Experiences in use of fresh water at Heavy Water Plants

    International Nuclear Information System (INIS)

    Heavy Water Board (HWB), an industrial organization under Department of Atomic Energy is primarily responsible for production of Heavy which is used as a Moderator and Coolant in the nuclear power reactors as well as research reactors. Two chemical exchange processes are being employed for industrial production of heavy water in India. H2S-H2O bi-thermal process requires approximately 35,000 tonnes of water to be processed to get one kg of nuclear grade water. Here surface water is the source of deuterium whereas H2S acts as deuterium carrier. Recently, HWB had indigenously developed front-end technology of ammonia-water exchange process for initial deuterium enrichment. In this process, water is the source of deuterium whereas NH3 acts as deuterium carrier. Moreover, these processes require elaborate heating and cooling arrangement in exchange process. Heating is accomplished through steam whereas cooling is effected through cooling and chilled water. This paper describes the measures adapted by HWB w.r.t. recycle and reuse and incorporating several process modifications and retrofitting in these plants to reduce consumption of this precious natural resource. As of now, cumulative water requirement for all six heavy water plants has drastically reduced from 2,50,000 m3 per day to the value of 61,000 m3 per day. (author)

  16. 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.

  17. The use of in situ gamma spectroscopy to study radionuclides contributing to dose rate at 540 MWe pressurised heavy water reactors

    International Nuclear Information System (INIS)

    Tarapur Atomic Power Station Unit-3 and 4 are twin reactors of 540 MWe capacity each. Unit-4 and Unit-3 operated for about 1030 and 910 effective full power days (EFPD) respectively. With the reactor operation, radiation field on reactor system equipments mainly on PHT system, Moderator system and spent fuel transfer system increases due to deposition of fission and activation product. These dose rates significantly contribute to the external exposure and stations collective dose in maintenance activities during reactor outages. In situ gamma spectroscopy has been successfully used at TAPS 3 and 4 operating nuclear facility to identify the radionuclide contributing to the dose rates for incorporating the corrective measures to control these sources and limit the exposures to ALARA. In situ gamma spectroscopy offers advantages over the traditional method of extracting a representative sample, transporting it to a laboratory, and then preparing the sample for counting. Some samples are physically difficult to obtain (material inside pipes, tanks, strainers, filters, very radioactive samples like resin beads, pressurized cover gases, heavy water sample collection). Since in situ spectroscopy is a non-contact process, and since the sample doesn't need to be physically extracted, these problems are minimized. In situ spectroscopy can give near-instantaneous results, and therefore allow prompt decisions to be made while the equipment is in the field. The availability of nuclide-specific information rather than just gross count or dose-rate information can allow better decisions to be made by the plant Health Physicist and plant management to define the optimum amount of personnel protection for the job. Reliable knowledge of exactly what radio nuclides are present, where they are located, will allow the job to be planned better. This better knowledge should lead to a safer operation, lower dose, lower risk of things going wrong, lower cost, and a quicker finish. Today

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

    International Nuclear Information System (INIS)

    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

  19. 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...

  20. Reactor water supplementing facility

    International Nuclear Information System (INIS)

    Condensates stored in a main condenser are introduced to a turbine-driven reactor feed water pump by way of a low pressure condensate pump, a condensate cleanup facility, a high pressure condensate pump and a low pressure feed water heater by condensate pipelines. The turbine driven feed water pump introduces feed water by way of a high pressure feed water heater to a reactor pressure vessel (RPV). Further, an auxiliary condensate pipeline having a booster pump and connected at one end to the main condenser is connected to the upstream of a motor-driven reactor feed water pump. Downstream of the turbine-driven feed water pump is connected to the downstream of the electromotive feed water pump. Then, when the condensate pump or a turbine-driven feed water pump should stop and if start of a stand-by pump is failed due to some or other reason, the motor-driven feed water pump and the booster pump are started based on a pump stop signal. With such procedures, coolants are supplied to RPV thereby enabling to ensure coolant level in the RPV. (I.N.)

  1. 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

  2. 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.

  3. 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)

  4. 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)

  5. 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

  6. ATLAS: a proposal for a precision heavy ion accelerator at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-02-01

    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 (5-25 MeV/A). By using the demonstrated potential of superconducting rf technology, beams of exceptional quality and flexibility can be obtained. The system is designed to provide beams with tandem-like energy resolution and ease of energy variation, the energy range is comparable to that of a approx. 50 MV tandem and, 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.

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

    International Nuclear Information System (INIS)

    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 (5-25 MeV/A). By using the demonstrated potential of superconducting rf technology, beams of exceptional quality and flexibility can be obtained. The system is designed to provide beams with tandem-like energy resolution and ease of energy variation, the energy range is comparable to that of a approx. 50 MV tandem and, 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. 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

  9. An Internal Tritium Concentration Analysis in Urine Samples as a Function of Submission Time after Airborne Tritium Intake at Korean Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    In pressurized heavy water reactors, workers who enter radiation controlled areas must submit their urine samples to health physicists after radiation work; these samples are then used to monitor internal radiation exposure from tritium intake. This procedure assumes that the samples submitted represent tritium concentration inside the body at equilibrium. According to both technical reports from the International Commission on Radiological Protection and experimental results from Canadian nuclear utilities, tritium inside the body generally reaches equilibrium concentration after approximately 2-3 hours of intake. In practice, urine samples can be submitted either before the 2 hours mark or after several hours of radiation work because of the numerous tasks that workers must perform and their frequent entries during nuclear power plant maintenance. In this paper, tritium concentration in workers' urine samples was measured as a function of time submitted after radiation work. Based on the measurement results, changes in the tritium concentration inside the body and its effect on internal dose assessment were then analyzed. As a result, it was found that tritium concentration reaches equilibrium concentration before the 2 hours mark for most workers' urine samples

  10. Methods of Containment Adopted for the EL4 Reactor and Projected Heavy-Water, Gas-Cooled Plants

    International Nuclear Information System (INIS)

    After a brief description of the plant, the paper explains the principles adopted for preventing the release of waste gas, from the EL4 reactor and refers to some of the difficulties associated with this type of containment. From the economic standpoint, the authors present the results of a comparative civil engineering study of pre-stressed concrete and steel shells for a projected 60 MW(e) power station, giving various values for accidental pressures. They demonstrate the influence of the stress values adopted. (author)

  11. Pre-service ultrasonic inspection of coolant channel rolled joints and surrounding areas in 540 MWe pressurised heavy water reactors

    International Nuclear Information System (INIS)

    To ensure safety of coolant channel rolled joints, in-situ, immersion ultrasonic testing was carried out to check presence of any unacceptable flaw. Twenty numbers of channels (≅ Square root of total channels in a reactor) covering central to peripheral region were selected for pre-service inspection. The calibration notches were made by EDM, both in axial and circumferential directions on OD and ID in adjacent and upset region. Sample rolled joint was cut open and freed from end fitting for making 2 % (0.086mm) depth artificial notches. Two numbers of inspection heads, housing two axial and two circumferential line focused ultrasonic probes were fabricated. In an annular Perspex cylindrical rod of 4 inch diameter, holes inclined at 27 deg were machined in axial and circumferential orientations. Immersion line focused probes of 10 MHz frequency were fitted in the drilled holes to generate 45 deg mode converted shear waves in the wall of pressure tube. No reportable flaw in the region adjacent to the rolled joint was observed in any channel of any of the two reactors. The coolant tube 11 to 1 'O' clock and between the two-rolled joint was also free from any discernible defect. (author)

  12. 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

  13. 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

  14. Design Details of the CGE Vertical Heavy-Water Pressure-Tube Reactor. Some Factors Influencing the Design

    International Nuclear Information System (INIS)

    The vertical HWR basic design criteria were to simplify existing designs to improve reliability and availability and to do this with existing proven technology. It was also a requirement to maintain access to operating areas at all times and to provide adequate shutdown shielding of the reactor to permit contact maintenance. The simplified fuel channel is characterized by; (a) Single-ended fuel changing on power using only one fuelling machine. (b) End-fitting closure which stores the seal. (c) Complete factory assembly and testing of the fuel channel before on-site installation. (This is permitted because of the use of only one end fitting.) (d) Fuel channel bolted to calandria permitting easy replacement of the assembly. This feature offers the possibility of retubing the reactor when better pressure-tube materials are developed (that is, the use of thinner pressure tubes to improve burnup which could be economically attractive if the price of uranium increases) . The simplified on-power refuelling is characterized by: (a) A simplified single fuelling machine using only mechanical drives and stops. (b) The capability of reshuffling the fuel into any desired order in the string by means of an accessible fuel handling mechanism. (c) The option of direct operator control of the fuelling machine. Improvements that have been made in reducing the building volume, dry vault concept and construction procedures by using the vertical design are described. In addition, the paper describes how the station can operate at base load to take advantage of low fuelling cost or how it may be used to control system frequency. The use of reactivity control mechanisms and steam by-pass to enable the power plant to operate with a continuously varying output over a wide range is described. (author)

  15. 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.

  16. Observations of the boiling process from a downward-facing torispherical surface: Confirmatory testing of the heavy water new production reactor flooded cavity design

    International Nuclear Information System (INIS)

    Reactor-scale ex-vessel boiling experiments were performed in the CYBL facility at Sandia National Laboratories. The boiling flow pattern outside the RPV bottom head shows a center pulsating region and an outer steady two-phase boundary layer region. The local heat transfer data can be correlated in terms of a modified Rohsenow correlation

  17. 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

  18. Purification by ion exchange resins of the heavy water of the reactors EL 1 and EL 2. A - the purifying process. Equipment and results

    International Nuclear Information System (INIS)

    The heavy water was purified by tapping off part of the moderator over a mixed bed of anion and cation exchangers. The heavy water leaving the columns has a resistivity reaching several-meg-ohms, which allows the resistivity of the moderator to be maintained between 105 and 106 ohms/cm. Two methods of deuteration of the ion exchangers are described, as well as the heavy water recuperation from resins charged with radioactive products. The influence of the purity of the water on the radiolytic dissociation is investigated. An interpretation of the variations in pH and of the formation of hydrogen peroxide is given. In addition the report contains a general description of the EL1 and EL2 purification installations. (author)

  19. The pressurized water reactor

    International Nuclear Information System (INIS)

    Pressurized water reactor technology has reached a maturity that has engendered a new surge of innovation, which in turn, has led to significant advances in the technology. These advances, characterized by bold thinking but conservative execution, are resulting in nuclear plant designs which offer significant performance and safety improvements. This paper describes the innovations which are being designed into mainstream PWR technology as well as the desings which are resulting from such innovations. (author)

  20. 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.

  1. 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).

  2. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  3. 'Experience with decommissioning of research and test reactors at Argonne National Laboratory'

    International Nuclear Information System (INIS)

    A large number of research reactors around the world have reached the end of their useful operational life. Many of these are kept in a controlled storage mode awaiting decontamination and decommissioning (D and D). At Argonne National Laboratory located near Chicago in the United States of America, significant experience has been gained in the D and D of research and test reactors. These experiences span the entire range of activities in D and D - from planning and characterization of the facilities to the eventual disposition of all waste. A multifaceted D nd D program has been in progress at the Argonne National Laboratory - East site for nearly a decade. The program consists of three elements: - D and D of nuclear facilities on the site that have reached the end of their useful life; - Development and demonstrations of technologies that help in safe and cost effective D and D; - Presentation of training courses in D and D practices. Nuclear reactor facilities have been constructed and operated at the ANL-E site since the earliest days of nuclear power. As a result, a number of these early reactors reached end-of-life long before reactors on other sites and were ready for D and D earlier. They presented an excellent set of test beds on which D and D practices and technologies could be demonstrated in environments that were similar to commercial reactors, but considerably less hazardous. As shown, four reactor facilities, plutonium contaminated glove boxes and hot cells, a cyclotron facility and assorted other nuclear related facilities have been decommissioned in this program. The overall cost of the program has been modest relative to the cost of comparable projects undertaken both in the U.S. and abroad. The safety record throughout the program was excellent. Complementing the actual operations, a set of D and D technologies are being developed. These include robotic methods of tool handling and operation, chemical and laser decontamination techniques, sensors

  4. Reactor water injection facility

    International Nuclear Information System (INIS)

    A steam turbine and an electric generator are connected by way of a speed convertor. The speed convertor is controlled so that the number of rotation of the electric generator is constant irrespective of the speed change of the steam turbine. A shaft coupler is disposed between the turbine and the electric generator or between the turbine and a water injection pump. With such a constitution, the steam turbine and the electric generator are connected by way of the speed convertor, and since the number of revolution of the electric generator is controlled to be constant, the change of the number of rotation of the turbine can be controlled irrespective of the change of the number of rotation of the electric generator. Accordingly, the flow rate of the injection water from the water injection pump to a reactor pressure vessel can be controlled freely thereby enabling to supply stable electric power. (T.M.)

  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 (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

  6. 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

  7. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    In the reactor operating with supercritical pressure and temperature part of the water flowing through the moderator tubes is deflected at the outlet and mixed with a residual partial flow of the coolant fed into the core as well as passed along the fuel rods in opposite direction. By special guiding of the flow downward through the guide tubes of the control rods insertion of the control rods is simplified because of reduced frictional forces. By this means it is also achieved to design less critical the control rod cooling with respect to flow rate control and operating behavior in case of a scram. (orig.)

  8. 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)

  9. 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

  10. 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

  11. A safety panel for heavy water facility

    International Nuclear Information System (INIS)

    The KUR Heavy Water Facility was renewed for preventive maintenance and for efficient utilization for medical research. This facility is monitored and operated by a new safety control panel installed by the irradiation room. Two types of the control system were employed. One is a traditional wired circuit system. The other is a supervisory, control and data acquisition (SCADA) system which runs on the Windows 3.1 composed of four personal computers connecting to a network called Arcnet. One of them is in a reactor control room and is also connected to Ethernet. (author)

  12. 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.)

  13. Heavy water extraction system

    International Nuclear Information System (INIS)

    In accordance with the present invention the tray for extracting liquid effluent from the humidifier section of the process is located in an intermediate position in the section, having a plurality of the trays of the section thereabove, with a balancing flow of heated water being admitted to the section beneath the extraction tray, to maintain substantially uniform loading of the trays in the humidifier section. The water being admitted in heat providing relation to the humidifier is divided into first and second portions, with the first portion admitted at the top of the humidifier section and the second portion substantially equal to the quantity of effluent extracted being admitted to the section beneath the extraction tray so as to maintain balance in the trays. The deuterium content of the gas, in passing through the trays above the extracting tray, is increased. The second water portion of the heater circuit is passed in heat exchanging relation with the effluent, to raise the temperature of the effluent prior to entry into the gas stripping portion of the plant wherein H2S gas is recovered for recycling. (author)

  14. 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

  15. 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)

  16. 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.

  17. 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

  18. 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

  19. 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) 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

  20. Use of CECE process for heavy water detritiation

    International Nuclear Information System (INIS)

    Full text: The experimental industrial plant has been built in PNPI for the development of the combined electrolysis catalytic exchange (CECE) technology for hydrogen isotope separation. The process uses a liquid phase catalytic exchange (LPCE) column and electrolysis cells to convert water to hydrogen. The plant has been in operation since 1995. In parallel with a development of CECE process for hydrogen isotope separation the plant is used for reprocessing tritium heavy water waste. Processing waste with the content of ∼47 % of heavy hydrogen and 108 Bq/kg of tritium, the plant produces 99.85-99.995% heavy water and deuterium gas for science and industry. Owing to industrial demands for heavy water with reduced tritium content, the plant was modified and additional equipment and procedures were put in place to operate in the detritiation mode. High detritiation factors have already been achieved during the initial heavy water detritiation test. In this test hydrogen catalytic burner is used as an upper reflux device. After prolonged operation campaigns it was decided to update the plant with an additional separation column connected with existing equipment. The main parameters of the new exchange column are the same as the old one: overall height of 7.5 m, inner diameter of 100 mm. The column is filled with alternating layers of hydrophobic catalyst developed by Mendeleev University and stainless steel spiral-prismatic packing. It consists of five separation sections connected through a distributor of liquid. The separation height of the column is 5.4 m. The new column and some additional equipment have been set in the plant so that tritiated deuterium enters the bottom of the column from electrolysis cells. Tritiated deuterium flowing up the column is purified from tritium and then enters the bottom of the second (old) column. The natural water flowing down in the second column is enriched in deuterium and is fed to the top of the first column. Deuterium

  1. 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

  2. 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.

  3. Chemistry in water reactors

    International Nuclear Information System (INIS)

    The international conference Chemistry in Water Reactors was arranged in Nice 24-27/04/1994 by the French Nuclear Energy Society. Examples of technical program areas were primary chemistry, operational experience, fundamental studies and new technology. Furthermore there were sessions about radiation field build-up, hydrogen chemistry, electro-chemistry, condensate polishing, decontamination and chemical cleaning. The conference gave the impression that there are some areas that are going to be more important than others during the next few years to come. Cladding integrity: Professor Ishigure from Japan emphasized that cladding integrity is a subject of great concern, especially with respect to waterside corrosion, deposition and release of crud. Chemistry control: The control of the iron/nickel concentration quotient seems to be not as important as previously considered. The future operation of a nuclear power plant is going to require a better control of the water chemistry than achievable today. One example of this is solubility control via regulation in BWR. Trends in USA: means an increasing use of hydrogen, minimization of SCC/IASCC, minimization of radiation fields by thorough chemistry control, guarding fuel integrity by minimization of cladding corrosion and minimization of flow assisted corrosion. Stellite replacement: The search for replacement materials will continue. Secondary side crevice chemistry: Modeling and practical studies are required to increase knowledge about the crevice chemistry and how it develops under plant operation conditions. Inhibitors: Inhibitors for IGSCC and IGA as well for the primary- (zinc) as for the secondary side (Ti) should be studied. The effects and mode of operation of the inhibitors should be documented. Chemical cleaning: of heat transfer surfaces will be an important subject. Prophylactic cleaning at regular intervals could be one mode of operation

  4. Water Cooled FBNR Nuclear Reactor

    International Nuclear Information System (INIS)

    A new era of nuclear energy is emerging through innovative nuclear reactors that are to satisfy the new philosophies and criteria that are developed by the INPRO program of the International Atomic Energy Agency (IAEA). The IAEA is establishing a new paradigm in relation to nuclear energy. The future reactors should meet the new standards in respect to safety, economy, non-proliferation, nuclear waste, and environmental impact. The Fixed Bed Nuclear Reactor (FBNR) is a small (70 MWe) nuclear reactor that meets all the established requirements. It is an inherently safe and passively cooled reactor that is fool proof against nuclear proliferation. It is simple in design and economic. It can serve as a dual purpose plant to produce simultaneously both electricity and desalinated water thus making it especially suitable to the needs of most of developing countries. FBNR is developed with the support of the IAEA under its program of Small Reactors Without On-Site Refuelling (SRWOSR). The FBNR reactor uses the pressurized water reactor (PWR) technology. It fulfills the objectives of design simplicity, inherent and passive safety, economy, standardization, shop fabrication, easy transportability and high availability. The inherent safety characteristic of the reactor dispenses with the need for containment; however, a simple underground containment is envisaged for the reactor in order to reduce any adverse visual impact. (author)

  5. Production of heavy water in Romania

    International Nuclear Information System (INIS)

    The production of heavy water from natural water is based on the water-sulfide hydrogen isotope exchange, in association with the isotope distillation process in vacuum. Isotope distillation in vacuum at 100 mm Hg is used to obtain a final heavy-water concentration of at least 99.8%. To ensure an independent operation of the heavy-water pilot factory, hydrogen sulfide and sodium sulfide production technologies have been developed and certified. The Romanian laboratory studies and the small scale development experience have been implemented in ROMAG DROBETA Tr.Severin, the Romanian Heavy Water Plant

  6. High Pressure Boiling Water Reactor

    International Nuclear Information System (INIS)

    Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are; 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340 oC (15 MPa) steam instead of ∼285 oC (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment -Well proved simple dry containment or wet containment can be used. (author)

  7. 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

  8. 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)

  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. Water simulation of sodium reactors

    International Nuclear Information System (INIS)

    The thermal hydraulic simulation of a large sodium reactor by a scaled water model is examined. The Richardson Number, friction coefficient and the Peclet Number can be closely matched with the water system at full power and the similarity is retained for buoyancy driven flows. The simulation of thermal-hydraulic conditions in a reactor vessel provided by a scaled water experiment is better than that by a scaled sodium test. Results from a correctly scaled water test can be tentatively extrapolated to a full size sodium system

  11. 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

  12. 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.

  13. Geneticaly modified flax for heavy metal phytoremediation

    Czech Academy of Sciences Publication Activity Database

    Najmanová, J.; Kotrba, P.; Macek, Tomáš; Macková, Martina

    Praha : VŠCHT, 2007 - (Macková, M.; Macek, T.; Demnerová, K.; Pazlar, V.; Nováková, M.), s. 167-168 ISBN 978-80-7080-026-3. [Symposium on Biosorption and Bioremediation /4./. Praha (CZ), 26.08.2007-30.08.2007] R&D Projects: GA MŠk 1M06030 Grant ostatní: GA MŠk(CZ) OC 117 Institutional research plan: CEZ:AV0Z40550506 Keywords : phytoremediation * heavy metal s * glutathione * flax * transformation Subject RIV: EI - Biotechnology ; Bionics

  14. General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel; Conception generale et principaux problemes d'un reacteur de puissance eau lourde-gaz contenu dans un caisson resistant

    Energy Technology Data Exchange (ETDEWEB)

    Roche, R.; Gaudez, J.C. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    In the framework of research carried out on a CO{sub 2}-cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [French] Dans le cadre des etudes d'un reacteur de puissance modere a l'eau lourde et refroidi-au gaz carbonique, la solution dite 'en caisson' consiste en une integration totale du coeur, du circuit primaire (echangeurs et soufflantes) et du dispositif de manutention du combustible a l'interieur d'un meme caisson etanche et resistant en beton precontraint. La disposition envisagee est verticale; le grenier de manutention est dispose au-dessus du coeur, les echangeurs en dessous. Cette solution, qui permet d'uniformiser les types de reacteurs moderes a l'eau lourde ou au graphite et refroidis par une circulation descendante de gaz carbonique presente, par rapport a la solution a tube de force, des avantages et des inconvenients qui sont analyses dans cette etude. L'extrapolation pose, en particulier, des problemes specifiques a l'eau lourde (tels que son refroidissement, son epuration

  15. Safety system in a heavy water detritiation plant

    International Nuclear Information System (INIS)

    In a CANDU 6 type reactor a quantity of 55·1015Bq/year of tritium is generated, 95% being in the D2O moderator which can achieve a radioactivity of 2.5-3.5·1012Bq/kg. Tritium in heavy water contributes with 30-50% to the doses received by operation personnel and up to 20% to the radioactivity released in the environment. The large quantity of heavy water used in this type of reactors (500 tones) make storage very difficult, especially for environment. The extraction of tritium from tritiated heavy water of CANDU reactors solve the following problems: the radiation level in the operation area, the costs of maintenance and repair reduction due to reduction of personnel protection measures, the increase of NPP utilisation factor by shutdown time reduction for maintenance and repair, use the extracted tritium for fusion reactors and not for the last, lower costs and risk for storage heavy water waste. Heavy water detritiation methods, which currently are used in the industrial or experimental plant, are based on catalytic isotope exchange or electrolysis followed cryogenic distillation or permeation. The technology developed at Institute of Cryogenics and Isotope Separation is based upon catalytic exchange between tritiated water and deuterium, followed by cryogenic distillation of hydrogen isotopes. The nature of the fluids that are processed in detritiation requires the operation of the plant in safety conditions. The paper presents the safety system solution chose in order to solve this task, as well as a simulation of an incident and safety system response. The application software is using LabView platform that is specialised on control and factory automation applications. (author)

  16. Sustainability of light water reactor fuel cycles

    International Nuclear Information System (INIS)

    This paper compares the sustainability of two light water reactor, LWR, fuel cycles: the once-through UOX (low-enriched uranium oxide) cycle and the twice-through MOX (Mixed Uranium-Plutonium Oxide) cycle (increasing the input efficiency of available uranium) by assessing their probable long-term competitiveness. With the retirement of diffusion enrichment facilities, enrichment prices have declined by one-third since 2009 and are likely to remain below $100-kgSWU for the foreseeable future. Here, initial uranium prices are set at $90/kgU and reprocessing costs at $2500 per kilogram of heavy-metal throughput, representative of “new-build” costs for reprocessing facilities. Substantial reprocessing cost reductions must be achieved if MOX is to be competitive, i.e., if it is to improve the sustainability of the LWR. However, results indicate that preserving the MOX alternative for spent fuel management later in this century has a large present value under several sets of assumptions regarding uranium price increases and reprocessing cost decreases. - Highlights: • We compare two nuclear fuel cycles: uranium versus reprocessed plutonium (mixed oxide, MOX). • We modify assumptions in MIT, The Future of the Nuclear Fuel Cycle (2011). • New reprocessing facilities are more expensive and uranium enrichment prices are lower than previously assumed, decreasing MOX competitiveness. • The MOX cycle option value could be large, but depends on uranium prices and reprocessing costs. • R and D should focus on reducing reprocessing facility costs before implementing the MOX fuel cycle

  17. Direct radiochemical monitoring of the coolant samples for fission gases. A new method for detecting fuel cladding failure in pressurised heavy water reactors

    International Nuclear Information System (INIS)

    The presence of radioactivity in the primary heat transport system of PHWR to a large extent depends upon 'Fuel Performance'. Any damage to the integrity of fuel cladding will lead to the release of gaseous and particulate fission products into the surrounding coolant medium and therefore for the safe operation of PHWR it is very essential to keep track of fuel performance. Though radioiodines and delayed neutron count rates give clue towards the condition of fuel clad integrity, they are found to be ambiguous in giving an assessment of defects which may develop in the fuel cladding. Hence, an attempt has been made for the first time to directly measure the chemically inert rare gas fission products in the heavy water of the primary heat transport system (PHT) and is presented in this paper

  18. Discovery of Interstellar Heavy Water

    Science.gov (United States)

    Butner, H. M.; Charnley, S. B.; Ceccarelli, C.; Rodgers, S. D.; Pardo, J. R.; Parise, B.; Cernicharo, J.; Davis, G. R.

    2007-04-01

    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 110-101 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 determine a total column density of N(D2O) of 1.0×1013 cm-2 and a fractional abundance of D2O/H2=1.7×10-10. The derived column density ratios for IRAS 16293-2422 are D2O/HDO=1.7×10-3 and D2O/H2O=5×10-5 for the hot corino gas. Steady state models of water ice formation, either in the gas phase or on grains, predict D2O/HDO ratios that are about 4 times larger than that derived from our observations. For water formation on grain surfaces to be a viable explanation, a larger H2O abundance than that measured in IRAS 16293-2422 is required. Alternatively, the observed D2O/HDO ratio could be indicative of gas-phase water chemistry prior to a chemical steady state being attained, such as would have occurred during the formation of this source. Future observations with the Herschel Space Observatory satellite will be important for settling this issue.

  19. Organically modified clay removes oil from water

    International Nuclear Information System (INIS)

    When bentonite or other clays and zeolites are modified with quaternary amines, they become organophilic. Such modified bentonites are used to remove mechanically emulsified oil and grease, and other sparingly soluble organics. If the organoclay is granulated, it is placed into a liquid phase carbon filter vessel to remove FOG's and chlorinated hydrocarbons. In this application the clay is mixed with anthrazite to prevent early plugging of the filter by oil or grease droplets. In batch systems a powered organoclay is employed. Types of oil found in water can include fats, lubricants, cutting fluids, heavy hydrocarbons such as tars, grease, crude oil, diesel oils; and light hydrocarbons such as kerosene, jet fuel, and gasoline

  20. Light-water nuclear reactors

    International Nuclear Information System (INIS)

    This work gives basic information on light-water reactors which is advanced enough for the reader to become familiar with the essential objectives and aspects of their design, their operation and their insertion in the industrial, economic and human environment. In view of the capital role of electric energy in the modern economy a significant place is given to electron-nuclear power stations, particularly those of the type adopted for the French programme. The work includes sixteen chapters. The first chapter relates the history and presents the various applications of light water reactors. The second refers to the general elementary knowledge of reactor physics. The third chapter deals with the high power light-water nuclear power station and thereby introduces the ensuing chapters which, up to and including chapter 13, are devoted to the components and the various aspects of the operation of power stations, in particular safety and the relationship with the environment. Chapter 14 provides information on the reactors adapted to applications other than the generation of electricity on an industrial scale. Chapter 15 shows the extent of the industrial effort devoted to light-water reactors and chapter 16 indicates the paths along which the present work is preparing the future of these reactors. The various chapters have been written to allow for separate consultation. An index of the main technical terms and a bibliography complete the work

  1. 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.

  2. Heavy water production by alkaline water electrolysis

    International Nuclear Information System (INIS)

    Several heavy water isotope production processes are reported in literature. Water electrolysis in combination with catalytic exchange CECE process is considered as a futuristic process to increase the throughput and reduce the cryogenic distillation load but the application is limited due to the high cost of electricity. Any improvement in the efficiency of electrolyzers would make this process more attractive. The efficiency of alkaline water electrolysis is governed by various phenomena such as activation polarization, ohmic polarization and concentration polarization in the cell. A systematic study on the effect of these factors can lead to methods for improving the efficiency of the electrolyzer. A bipolar and compact type arrangement of the alkaline water electrolyzer leads to increased efficiency and reduced inventory in comparison to uni-polar tank type electrolyzers. The bipolar type arrangement is formed when a number of single cells are stacked together. Although a few experimental studies have been reported in the open literature, CFD simulation of a bipolar compact alkaline water electrolyzer with porous electrodes is not readily available.The principal aim of this study is to simulate the characteristics of a single cell compact electrolyzer unit. The simulation can be used to predict the Voltage-Current Density (V-I) characteristics, which is a measure of the efficiency of the process.The model equations were solved using COMSOL multi-physics software. The simulated V-I characteristic is compared with the experimental data

  3. 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.)

  4. 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.)

  5. 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

  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. 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 ...

  8. Review of the use and state of development of the various reactor types

    International Nuclear Information System (INIS)

    The report gives a review of the reactor types being of importance from today's point of view for use as stationary power reactors. These are heavy water reactors, light water reactors (pressurized water reactor, Soviet pressurized water reactor, Soviet light-water-graphite reactors, boiling water reactors), gas-cooled reactors (gas-graphite reactors, high temperature reactors), and fast breeder reactors. (HJ)

  9. European supercritical water cooled reactor

    International Nuclear Information System (INIS)

    Highlights: → The HPLWR reactor design is an example of a supercritical water cooled reactor. → Cladding material tests have started but materials are not yet satisfactory. → Numerical heat transfer predictions are promising but need further validation. → The research project is most suited for nuclear education and training. - Abstract: The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 oC average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers

  10. 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

  11. Design of a single cavity harmonic buncher for Argonne's low beta heavy ion linac

    International Nuclear Information System (INIS)

    A compact harmonic buncher cavity has been designed for Argonne's low beta Xe/sup +1/ linac. The resonant sections are nested together to save beam line space. The cavity is approximately 1.5 m in height and at its maximum 0.4 m in outside diameter. It takes up only 0.22 m of beam line space. The computer program ''Superfish'' and a transmission line model of the resonator were used to calculate its properties. Considering the complicated geometry, excellent agreement (within 1%) was achieved. 8 refs

  12. 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.

  13. 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

  14. 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.)

  15. 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.)

  16. Radiological safety aspects: draining of heavy water during preparatory phase of proposed decommissioning of Cirus

    International Nuclear Information System (INIS)

    CIRUS reactor, the nucleus of nuclear programme of India was a 40 MWth research reactor, commissioned in July 1960. After five decades of valuable service to the nation, reactor was permanently shut down on 31st December 2010. The reactor used natural uranium as fuel and was cooled by demineralized water in a closed loop. The Heavy water (D2O) was used as moderator and helium gas was used as cover gas in the reactor vessel. Moderator system removed about 5% heat of the reactor power. This heat in turn was removed by sea water in heat exchangers. System tritium activity before permanent shutdown of Cirus reactor was 5.0 Ci/l. Tritium is the main radioactive isotope present in the moderator system which can cause internal hazards. During preparatory phase of decommissioning activities heavy water was drained from reactor vessel, heavy water heat exchangers, ion exchangers, different loops and instruments. Drained heavy water was transferred to storage tank no. 1 for interim storage. On the spot cold finger sampling was done to monitor tritium activity in air during and after each job related to handling of heavy water. Maximum tritium activity measured in air was 213 DAC. Internal dose due to tritium for all the jobs was 2.65 pmSv. This paper gives details about the radiological safe procedures followed, monitoring of tritium activity in air and internal dose assessment due to tritium during draining of heavy water from reactor moderator system. Implementation of operational practices and radiological safety coverage provided during the job resulted in low internal dose due to tritium and negligible rise in stack releases. (author)

  17. Status of control assembly materials in Indian water reactors

    International Nuclear Information System (INIS)

    India's present operating water cooled power reactors comprise boiling water reactors of Tarapur Atomic Power Station (TAPS) and pressurized heavy water reactors (PHWRs) at Kota (RAPS), Kalpakkam (MAPS), Narora (NAPS) and Kakrapara (KAPS). Boiling water reactors of TAPS use boron carbide control blades for control of power as well as for shut down (scram). PHWRs use boron steel and cobalt absorber rods for power control and Cd sandwiched shut off rods (primary shut down system) and liquid poison rods (secondary shut down system) for shut down. In TAPS, Gadolinium rods (burnable poison rods) are also incorporated in fuel assembly for flux flattening. Boron carbide control blades and Gadolinium rods for TAPS, cobalt absorber rods and shut down assemblies for PHWRs are fabricated indigenously. Considerable development work was carried out for evolving material specifications, component and assembly drawings, and fabrication processes. Details of various control and shut off assemblies being fabricated currently are highlighted in the paper. (author)

  18. 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.)

  19. Application of fuzzy logic control system for reactor feed-water control

    International Nuclear Information System (INIS)

    The successful actual application of a fuzzy logic control system to the a nuclear Fugen nuclear power reactor is described. Fugen is a heavy-water moderated, light-water cooled reactor. The introduction of fuzzy logic control system has enabled operators to control the steam drum water level more effectively in comparison to a conventional proportional-integral (PI) control system

  20. 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.)

  1. 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.

  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. 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.

  4. Safety and hazard control in Heavy Water Plants

    International Nuclear Information System (INIS)

    Heavy Water Plants are constituent units of Heavy Water Board under Department of Atomic Energy. Heavy water, a compound of deuterium (an isotope of hydrogen) and oxygen is used as a moderator and coolant in Pressurised Heavy Water based nuclear reactors. Heavy Water Plants at Baroda, Tuticorin, Thal and Hazira employ Mono-thermal Ammonia Hydrogen Chemical Exchange Process for production of heavy water. Raw materials used in this process are Deuterium bearing hydrogen in the form of synthesis gas (a mixture of hydrogen and nitrogen in the ratio of 3:1) produced in the adjacent fertilizer plants. HWPs at Kota and Manuguru employ bi thermal Water-Hydrogen Sulphide Chemical Exchange Process. HWPs handle hazardous chemicals such as Hydrogen, Hydrogen sulphide, Ammonia, Naphtha, Hexane, Potassium, Potassium Amide etc. The operating pressure and temperature are varying from vacuum to 260 kg/cm2(g) and -185 deg C to 900 deg C respectively. Our commitment towards safety and health of human resources, materials and environment have paid rich dividends to HWPs. HWPs have achieved many records and received many awards including 4414 days operation (twelve years of operation) continuously without any lost time injury, several awards including Industrial Safety Award from AERB (Atomic Energy Regulatory Board), highest national award in safety among chemical industries - Sarvashresta Suraksha Puraskar India consecutively two years and Shresta Suraksha Puraskar for five years from NSC (National Safety Council), India. The high standard of performance could be achieved mainly due to self discipline, skilled operation of the plants with improved work culture, Quality Management System ISO 9001, Environment Management System ISO 14001 and Occupational Health and Safety Management System IS 18001, adopted in HWPs besides well defined responsibility and authority in HWB and HWPs

  5. 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)

  6. Water cooled FBNR nuclear reactor

    International Nuclear Information System (INIS)

    Full text: The world with its increasing population and the desire for a more equitable and higher standard of living, is in the search for energy that is abundant and does not contribute to the problem of global warming. The answer to this is a new paradigm in nuclear energy; i.e., through the innovative nuclear reactors that meet the IAEA's INPRO philosophies and criteria that will guarantee the generation of safe and clean energy. The emerging countries to nuclear energy that are not in hurry for energy and look into the future are looking into the participation in the development of such innovative nuclear reactors. They can start developing the non-nuclear components of such reactors in parallel with creating the nuclear infra-structures according to the guidelines of the IAEA suggested in its milestones document. In this way, they can benefit from numerous advantages that the development of a high technology can bring to their countries be it scientific, technological, economic or political. A solution to the present world economic crisis is investing in such projects that contribute to the real economy rather than speculative economy. This will help both local and world economy. One such innovative nuclear reactor is the FBNR that is being developed with the support of the IAEA in its program of Small Reactors Without On-site Refuelling. It is a small (70 MWe) reactor with simple design based on the proven PWR technology (www.sefidvash.net/fbnr). The simplicity in design and the world wide existence of water reactor technology, makes it a near term project compared to other future reactors. Small reactors are most adequate for both the developing and developed countries. They require low capital investment, and can be deployed gradually as energy demand calls for. The generation of energy at the local of consumption avoids high cost of energy transmission. The paradigm of economy of scale does not apply to the FBNR as it is a small reactor by its nature. The

  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. Thermal analysis of LEU modified Cintichem target irradiated in TRIGA reactor

    International Nuclear Information System (INIS)

    Actions conceived during last years at international level for conversion of Molybdenum fabrication process from HEU to LEU targets utilization created opportunities for INR to get access to information and participating to international discussions under IAEA auspices. Concrete steps for developing fission Molybdenum technology were facilitated. Institute of Nuclear Research bringing together a number of conditions like suitable irradiation possibilities, direct communication between reactor and hot cell facility, handling capacity of high radioactive sources, and simultaneously the existence of an expanding internal market, decided to undertake the necessary steps in order to produce fission molybdenum. Over the course of last years of efforts in this direction we developed the steps for fission Molybdenum technology development based on modified Cintichem process in accordance with the Argonne National Laboratory proved methodology. Progress made by INR to heat transfer computations of annular target using is presented. An advanced thermal-hydraulic analysis was performed to estimate the heat removal capability for an enriched uranium (LEU) foil annular target irradiated in TRIGA reactor core. As a result, the present analysis provides an upper limit estimate of the LEU-foil and external target surface temperatures during irradiation in TRIGA 14 MW reactor. (authors)

  9. Innovative Pressure Tube Light Water Reactor with Variable Moderator Control

    International Nuclear Information System (INIS)

    The features of a reactor based on multiple pressure tubes, rather than a single pressure vessel, provide the reactor with considerable flexibility for continuous design improvements and developments. This paper presents the development of innovative pressure tube light water reactor, which has the ability to advance the current pressure tubes reactors. The proposed design is aimed to simplify the pressure tubes reactors by: - replacing heavy water by a light water as a coolant and moderator, - adopting batch refueling instead of on-line refueling. Furthermore, the design is based on proven technologies, existing fuel and structure materials. Therefore, it is reasonable to expect significant capital cost savings, short licensing and introduction period of the proposed concept into the power production grid. The basic novelty of the proposed design is based on an idea of variable moderator content in the core and 'breed and burn' mode of operation. Both concepts were extensively investigated and reported in the past (2) (3) (4). In order to evaluate a practical reactor design build on proven technology, several features of the advanced CANDU reactor (ACR-1000) were adopted. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The proposed design is basically pressure tube light water reactor with variable moderator Control (PTVM LWR). This paper presents a detailed description of the PTVM core design and demonstrates the reactivity control and the 'breed and burn' mode of operation, which are implemented by the variation of the moderator in the core, from a

  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. Critical heat flux measurements in rod bundles using light water and heavy water as coolant

    International Nuclear Information System (INIS)

    A series of Critical Heat Flux (CHF), subchannel mixing and pressure drop tests were performed on a full scale simulated 37 element fuel bundle at the Heat Transfer Research Facility of Columbia University for Siemens/Kraftwerk Union Ag. of Germany. The experimental program consisted of thermal hydraulic testing of a full scale 37 element fuel bundle 3000 mm long to determine two phase pressure drop and CHF characteristics. The test were performed on the simulated fuel bundle in a vertical circular test housing with an outer diameter of 108.25 mm. The axial power distribution was uniform, while the radial power distribution was non-uniform with a heat flux depression of about 35%. The bundle geometry was maintained by spacer grids. In this experimental program, 92 CHF runs were performed using light water (H2O) as the coolant, and 22 CHF runs were performed using heavy water (D2O) as the coolant to reduce the risk of heavy water loss. The CHF tests covered the following parameter ranges: pressure from 70 to 150 bars in light water and at 100 and 115 bars in heavy water; mass velocities from 1000 to 5600 kg/m2s, and inlet temperatures from 200 to 320 degree C. The experiments carried out in heavy water were in essence a duplication of the test conditions of the ones obtained in light water, facilitating a direct comparison of the results using both fluids. Furthermore, an evaluation of the experiments using the subchannel analysis method was made with the results being presented. Analysis shows that CHF with heavy water as a coolant can be predicted with sufficient accuracy by applying Ahmad's scaling laws for fluid -to fluid modeling. Therefore, CHF correlations developed and verified for light water reactors can be applied to design heavy water cooled reactor cores

  12. 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)

  13. 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

  14. 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.

  15. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    International Nuclear Information System (INIS)

    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

  16. The utilisation of thorium fuel in a generation 4 light water reactor design

    International Nuclear Information System (INIS)

    During the last several years the Department of Energy has sponsored research at Purdue University on advanced reactor designs under the Nuclear Energy Research Initiative (NERI) programme. This work has involved research in 'Generation IV' advanced reactor designs such as a high conversion boiling water reactor, as well research in advanced fuel designs such a metal matrix 'dispersion' fuel. The unifying theme of this research has been to take advantage of the numerous benefits of the thorium fuel cycle. The Purdue research has been performed in collaboration with Argonne and Brookhaven National Laboratories for the dispersion fuel research and the high conversion reactor research, respectively. The primary contribution to both research efforts from Purdue has been on neutronics design and analysis. This paper will focus on the neutronics design and analysis of the high conversion boiling water reactor. (author)

  17. Fatigue properties in reactor water

    International Nuclear Information System (INIS)

    Results of research projects on the effect of reactor water environment on the fatigue properties of the pressure boundary materials have been reported in the literature for the last 20 years. These include projects to investigate effect on crack growth along with projects on crack initiation through studies of the S-N fatigue properties. Much has been learned in these studies about how the environment can influence the fatigue life of a material in an environment. Much still needs to be done, however, before this information can be applied to the analyses of actual structures. This presentation will present an overview of what is known concerning the interactions of the reactor environment with the pressure boundary materials that affect the fatigue properties of the material, a summary of current activities within the PVRC and a discussion of the work needed before the environmental effects can be effectively considered in the analysis of structures

  18. Heavy metal adsorption by modified oak sawdust: Thermodynamics and kinetics

    International Nuclear Information System (INIS)

    This paper describes the adsorption of heavy metal ions from aqueous solutions by oak (Quercus coccifera) sawdust modified by means of HCl treatment. Our study tested the removal of three heavy metals: Cu, Ni, and Cr. The optimum shaking speed, adsorbent mass, contact time, and pH were determined, and adsorption isotherms were obtained using concentrations of the metal ions ranging from 0.1 to 100 mg L-1. The adsorption process follows pseudo-second-order reaction kinetics, as well as Langmuir and D-R adsorption isotherms. The paper discusses the thermodynamic parameters of the adsorption (the Gibbs free energy, entropy, and enthalpy). Our results demonstrate that the adsorption process was spontaneous and endothermic under natural conditions. The maximum removal efficiencies were 93% for Cu(II) at pH 4, 82% for Ni(II) at pH 8, and 84% for Cr(VI) at pH 3

  19. Future trends in heavy water production

    International Nuclear Information System (INIS)

    World heavy water production has spanned nearly fifty years and, for much of that period, the commodity was often in short supply, but that situation has changed, at least in Canada. There are now adequate reserves of heavy water and sufficient installed production capacity to service Canadian domestic and export demands for the next ten years or beyond. More than 90 percent of the world's inventory of heavy water has been produced by the GS process but this may not be the method that is chosen when the time comes to expand heavy water production again. Other countries, such as India and Argentina, have already chosen ammonia-hydrogen exchange as an alternative technology for part of their domestic production programs. Despite the present surplus of heavy water, research and development of new technologies is very active, particularly in Canada and Japan, because it is recognized that there are still attractive opportunities for future production by processes that are both less expensive and environmentally more acceptable, than either the demonstrated GS process or ammonia-hydrogen alternative. This paper describes the prospects for some of these new processes, contrasts them with the present established methods and assesses the probable impact on the future supply situation

  20. Design of a single-cavity harmonic buncher for Argonne's low-beta heavy-ion linac

    International Nuclear Information System (INIS)

    In order to increase the beam capture efficiency of Argonne's heavy-ion linac to over 70%, a single cavity harmonic buncher has been designed as a replacement for the existing fundamental frequency buncher. Because the beam line space between the 1.5-MeV Xe+1 preaccelerator and first accelerating cavity is at a premium, especially in the tunnel area near the preaccelerator, a single-cavity design was undertaken. In addition, to further conserve access space, the cavity was designed to fit directly beneath the beam line. The cavity is designed to resonate at the fundamental linac frequency of 12.5 MHz and its first harmonic, 25 MHz. This was accomplished by nesting the 25-MHz resonant section inside the larger 12.5-MHz resonant section. Both sections are heavily capacitively loaded, folded coaxial lines with two 0.008-m accelerating gaps per section. The cavity was designed using a transmission line model taking account of the capacitances of each discontinuity and by use of the RF cavity computer program Superfish. The transit time factor for the cavity gaps was calculated using the computer program Poisson and are 0.44 for the 25-MHz section and 0.70 for the 12.5 MHz section. The transit time factors are poor because of the large linac aperture of 0.049 m

  1. The Shippingport Pressurized Water Reactor and Light Water Breeder Reactor

    International Nuclear Information System (INIS)

    This report discusses the Shippingport Atomic Power Station, located in Shippingport, Pennsylvania, which was the first large-scale nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. A program was started in 1953 at the Bettis Laboratory to confirm the practical application of nuclear power for large-scale electric power generation. It led to the development of zirconium alloy (Zircaloy) clad fuel element containing bulk actinide oxide ceramics (UO2, ThO2, ThO2 -- UO2, ZrO2 -- UO2) as nuclear reactor fuels. The program provided much of the technology being used for design and operation of the commercial, central-station nuclear power plants now in use. The Shippingport Pressurized Water Reactor (PWR) began initial power operation on December 18, 1957, and was a reliable electric power producer until February 1974. In 1965, subsequent to the successful operation of the Shippingport PWR (UO2, ZrO2 -- UO2 fuels), the Bettis Laboratory undertook a research and development program to design and build a Light Water Breeder Reactor (LWBR) core for operation in the Shippingport Station. Thorium was the fertile fuel in the LWBR core and was the base oxide for ThO2 and ThO2 -- UO2 fuel pellets. The LWBR core was installed in the pressure vessel of the original Shippingport PWR as its last core before decommissioning. The LWBR core started operation in the Shippingport Station in the autumn of 1977 and finished routine power operation on October 1, 1982. Successful LWBR power operation to over 160% of design lifetime demonstrated the performance capability of the core for both base-load and swing-load operation. Postirradiation examinations confirmed breeding and successful performance of the fuel system

  2. 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

  3. Modified silicates applied in adsorption of heavy metal

    International Nuclear Information System (INIS)

    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)

  4. 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)

  5. 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.

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

    International Nuclear Information System (INIS)

    This report documents the rate constants and associated activation energies for the reactions of the primary radical species, eaq-, ·OD and ·D, which are formed during the radiolysis of heavy water within the temperature range 20 to 250 oC. 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)

  7. Fabrication of fuel for water reactors

    International Nuclear Information System (INIS)

    Nuclear Fuel Complex (NFC) established at Hyderabad in 1972, produces zircaloy clad uranium dioxide fuels for pressurised heavy water reactors (PHWRs) at Kota (Rajasthan), Kalpakkam (Tamilnadu), Narora (U.P.) etc., and boiling water reactors (BWRs) at Tarapur (Maharashtra). The experience gained at BARC in the production of zircaloy clad natural uranium oxide fuel for the initial core of the first PHWR at Kota, has been translated and utilised for establishing higher capacity fabrication plant at NFC. Natural uranium dioxide fuel pellets are produced starting with indigenous magnesium di-uranate concentrate from Uranium Corporation of India Ltd., Jaduguda (UCIL), a unit of the Department of Atomic Energy (DAE); and enriched uranium dioxide pellets from imported uranium hexafluoride and then fabricated into fuel elements and assemblies using zircaloy fuel tubes and components produced at NFC. For PHWR fuels the yellow cake, MDU, is purified through solvent extraction process and converted to ceramic grade uranium oxide powder via ADU route which is then pelletised and the pellets are loaded in zircaloy tubes which in turn are resistance welded with end caps to form elements. 19 elements with strip appendages are assembled into bundles with zircaloy end plates and the fuel bundles supplied to the various reactor sites. For BWR fuels the imported enriched uranium hexafluoride, in designated enrichments, is processed through ADU route to produce ceramic grade uranium oxide powder which is pelletised and the pellets loaded in autoclaved zircaloy tubes and the loaded tubes in turn TIG welded with end plugs to form elements. 36 elements of different U235 enrichments (including two elements of UO2-Gd2O3) are assembled in 6x6 array with stainless steel tie plates at the end, with the intermediate spacing maintained by zircaloy spacer grids. (author) 2 figs

  8. 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)

  9. 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.

  10. 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

  11. Analysis of water cooled reactors stability

    International Nuclear Information System (INIS)

    A model for stability analysis of non-boiling water cooled nuclear system is developed. The model is based on linear reactor kinetics and space averaged heat transfer in reactor and heat-exchanger. The transfer functions are defined and the analysis was applied to nuclear reactor RA at 'Boris Kidric' Institute - Vinca. (author)

  12. 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.

  13. Environmental assessment related to the decontamination and decommissioning of the Argonne National Laboratory CP-5 research reactor

    International Nuclear Information System (INIS)

    Five alternatives for the decontamination and decommissioning of the Argonne National Labortory CP-5 research reactor are considered. Results of this study on environmental changes and impacts due to the action indicate that there will be no adverse impact on land use; decommissioning of the facility will release about 1.2 ha (3 acres) of a previously restricted area for unrestricted use, whereas radioactive-waste burial will occupy only an estimated 0.03 ha (0.07 acre. Some of the biotic habitat, vegetation, and animal life of the 1.2-ha (3-acre) waste-storage yard will be disturbed or destroyed during decontamination of the yard. The impact will be negligible in terms of the local ecosystem. There will be minimal socioeconomic impact on the area. Radiological impacts on the population from nonaccidental releases of the radionuclides 3H, 60Co, 55Fe, and 63Ni will include a dose commitment possibly as high as 0.19 mrem to the lungs of an individual working onsite and located about 100 m (300 ft) to the northeast of the reactor building. The cumulative dose to the population within an 80-km (50-mi) radius is 8.33 person-rem; this is about 10-5 of the annual natural-background dose for this area. The risks of significant radiological impacts on the population from accidents of natural catastrophies at the reactor site are extremely small. A cumulative occupational dose of about 21 person-rem will be received by the work force of up to about 50 persons participating in the dismantling activities. Population doses during the transportation of reactor scrap and wastes from dismantlement will be about 50% of the cumulative population dose within 80 km of the site. A cumulative occupational dose of about 24 person-rem could be received by the drivers of the transport trucks shipping the radioactive wastes to Richland, Washington

  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. High performance light water reactor

    International Nuclear Information System (INIS)

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

  16. Problems of hydrogen - water vapor - inert gas mixture use in heavy liquid metal coolant technology

    International Nuclear Information System (INIS)

    The reasons of slag deposit formation in circulation circuits with heavy liquid metal coolants, which can cause reactor core blockage, are considered. To prevent formation of deposits hydrogen purification of coolant and surfaces of circulation circuit is used. It consists in introduction of gaseous mixtures hydrogen - water vapor - rare gas (argon or helium) directly into coolant flow. The principle scheme of hydrogen purification and the processes occurring during it are under consideration. Measures which make it completely impossible to overlap of the flow cross section of reactor core, steam generators, pumps and other equipment by lead oxides in reactor facilities with heavy liquid metal coolants are listed

  17. 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

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

    International Nuclear Information System (INIS)

    Fast reactors cooled by lead or lead-bismuth alloy offer new interesting fuel cycle and fuel management options by virtue of the superb neutronics and safety features of these heavy liquid metal (HLM) coolants. One option is once-for-life cores having relatively low power density. These cores are fueled in the factory; there is no refueling or fuel shuffling on site. A second option is very long-life cores being made of a fissioning zone and a natural uranium blanket zone. The fissioning zone very slowly drifts toward the blanket. A third option is multirecycling of light water reactor (LWR) discharged fuel without partitioning of transuranics (TRUs) in fuel-self-sustaining reactors. LWR spent fuel could provide the initial fuel loading after extracting fission products and ∼90% of its uranium. The makeup fuel is natural or depleted uranium. A fourth option is the high-burnup once-through fuel cycle using natural or depleted uranium feed. The initial fuel loading of this reactor is a mixture of enriched and natural uranium. The natural uranium utilization is 10 to 20 times higher than that of a once-through LWR. A fifth option is transmutation of TRUs from LWRs using critical HLM-cooled reactors; such reactors could be designed to have the same high actinide burning capability of accelerator-driven systems and have comparable safety, but at a substantially lower cost. These novel reactor designs and fuel management options are hereby reviewed

  19. On the domestically-made heavy forging for reactor pressure vessels of PWR nuclear power plant

    International Nuclear Information System (INIS)

    The present situation of the foreign heavy forgings for nuclear reactor pressure vessels and the heavy forgings condition which is used for the Qinshan 300MWe nuclear power plant are described. Some opinions of domestic products is proposed

  20. 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

  1. 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

  2. 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

  3. Validation of the modified ATHLET code with the natural convection test of the PHENIX reactor

    International Nuclear Information System (INIS)

    Highlights: • Modification of system code ATHLET for Sodium-cooled Fast Reactors application. • Development of a properties package as well as a heat transfer package for sodium. • Validation of the modified code with the PHENIX reactor ultimate natural convection test. - Abstract: This paper presents the modification, validation and application of the system code ATHLET for Sodium-cooled Fast Reactors. The ATHLET code is developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) in Germany for Light Water Reactors application. The code structure is modified that it can be easily extended to different fluids. For the present analysis of SFR, a sodium property package as well as heat transfer correlations for sodium are implemented into the code. To evaluate the feasibility of the modified code, the PHENIX reactor, a SFR operated by French Alternative Energies and Atomic Energy Commission (CEA) from 1973 to 2009, is modeled. Two different modeling approaches of the hot and cold plenum of the PHENIX reactor are adopted, i.e. the one-dimensional representation with standard pipes, and the pseudo-three-dimensional representation with parallel channels connected via cross connections. Both models are used for simulation of the natural convection ultimate test scenario of the PHENIX reactor, and the results are compared with the measured data. Improvement approach for the pseudo-three-dimensional modeling method is proposed and realized. The results reveal advantage as well as limitation of the current models, and good applicability of the modified ATHLET code to sodium-cooled reactor systems

  4. Using MAGNA-IR 750 spectrometer in heavy water measurements

    International Nuclear Information System (INIS)

    Since ten years the Laboratory of Deuterium Metrology is operational and provides the calibration standard to the units throughout the country which produce and use heavy water. During this period interlaboratory comparisons were conducted with laboratories of similar character of Argentina, Canada, India and USA. In 1994 an indirect intercalibration with Canada to determine the concentration of heavy water introduced for the first loading of the Cernavoda Unit 1 reactor has been performed. In 1996 a new intercalibration has been made with the occasion of handing back the heavy water lent by the Canadian party. This work presents the measurement techniques and the practical results obtained in the Laboratory of Deuterium Metrology of INC-DTCI-ICIS Rm.Valcea. The analyses were carried out with the MAGNA-IR 750 spectrophotometer produced by NICOLET. This type of instruments is based on a interferometer and makes use of Fourier transform, what leads to improving the infrared spectrum quality and 'implicitly' to increase the accuracy of the results obtained up to value of 0.005% of the D2O mass. Also, a significant reduction of the time needed for making an analysis has been achieved

  5. 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

  6. Subcriticality Evaluation of AGN-201 Reactor Using Modified Neutron Source Multiplication Method

    International Nuclear Information System (INIS)

    One of the main issues in nuclear criticality safety is to measure subcriticality accurately at nuclear facility containing fissile materials. In order to verify the feasibility and safety of reactor, reactor physics test is performed in the commercial reactor. Among these test items, the measurement of control rod worth is taken most of period of reactor physics test. For that reason, the new methods have been introduced for subcriticality measurement to reduce the test period from the economic point of view : for example, pulse neutron method, neutron noise analysis method, Neutron Source Multiplication (NSM) method and so on. In 1980's, the research for subcriticality measurement methodology was performed about accelerator driven system, fast breeder reactor and critical experiment reactor. In this study, subcritcality is evaluated by modified NSM method. It is based on the conventional NSM method adding two correction processes: extraction of the fundamental mode from measuring neutron count rate data that contains not only fundamental mode but also higher modes in real situation and spatial corrections for perturbation induced by a reactivity addition in the distributions of the fundamental mode and a neutron importance field. In the previous studies, the verification of this method has been firstly performed for the subcriticality measurement of critical assembly of Kyoto University Critical Assembly (KUCA) at Kyoto University Research Reactor Institute in Japan. Recently subcriticality measurement study for the Pressurized Water Reactor (PWR) has been carried out. In the present study, the subcriticality was evaluated for Aerojet General Nucleonics (AGN)-201 reactor by the modified NSM method with two correction processes. The AGN-201 reactor is the graphite moderated homogeneous type research reactor and is used for reactor experiments such as critical mass approach, control rod calibration, measurement of neutron flux and so on. For subcriticality

  7. 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)

  8. 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.

  9. Sediment, water pollution indicators for heavy metals

    International Nuclear Information System (INIS)

    The complexity of an aquatic system requires consideration of its dynamics: spatial and temporal variations of physical, chemical and biological. Heavy metals have peculiar behavior in the aquatic system and may not be available in the waters, but on sediments.The sub-basin of the Sarandi stream is responsible for the contamination of Pampulha Lake. The Instituto Mineiro das Águas – IGAM - uses tool for monitoring the quality of surface water for developing strategies for conservation, restoration and rational use of water resources. So through the indices: IQA ( Indice de qualidade de águas) Index of water quality, and TC- toxic contamination, reduces conflicts, implements the disciplining of the environmental economy.This study determined the monitoring of sediment and water of Sarandi Stream, so in the samples collected during dry and rainy seasons (2007- 2008) were analyzed heavy metals (Cu, Cd, Cr, Co, Ni, Zn, Pb) and physical-chemical factors (conductivity, solids dissolved, temperature, turbidity). This allowed the determination of Hackanson factors of contamination and Muller Index geoaccumulation, indicating very high contamination in sediments regarding the elements Cr, Cu, and Cd, and high contamination for Pb, Zn, and Mn. The comparison with the indices of water quality- IQA (IGAM - 2006, 2007 and 2008), combined with exploratory data analysis and graphs of correlation between the variables indicated favorable conditions for metals contamination on water and sediment for these metals, besides allowing the identification of its source

  10. Safety system in a heavy water detritiation plant

    International Nuclear Information System (INIS)

    In the heavy water circuits of the CANDU reactor, tritium is generated through the following reactions: - neutron activation; - 235 U, 233 U and 239 Pu fission in rods; - 3 He decay. The equilibrium value, of about 30-50 Ci/kg, depends by the heavy water quantity which is used in common in primary circuit and moderator, at an increase rate by 4Ci/kg/year. As the tritium inventory in a CANDU reactor can be larger than 6 x 106 Ci (tritium oxide in heavy water circuits), it results that a detritiation plant is a necessity. Solving these problems means that a heavy water detritiation plant must built and linked to the moderator circuits of the CANDU type reactor. This plant can be assimilated as a nuclear facility, involving special regulation and safety systems, complying with the nuclear laws of Romania and international safety regulations, including IAEA Vienna specifications. Like any nuclear facility, a special safety system is provided, with special hardware and software that supervises the technological process and safety equipment. Conventional systems use a large number of equipment, very expensive, while the reliability and accuracy are basic demands. On the other hand, the systems become more complex solving demands like redundancy, failure of safety or diversity. Like a result, operation and maintenance become more complicated and more expensive. Solution for these problems is to develop a reliable and flexible on-line diagnosis system, comprising two computers and a small number of discrete equipment. Creating a safety display and analysis system that provides an overview of the plant safety status and prevents serious safety degradation, ensures an easy maintenance and operation. The system is flexible, easy to use and the improvements needed by any technological process experiment could be done in short time and at low costs. Such a system can replace a dedicated hardware and software for industrial processes, regarding especially the experimental

  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. 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

  13. Pressurized water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator

  14. Dual pressurized light water reactor producing 2000 M We

    International Nuclear Information System (INIS)

    The dual unit optimizer 2000 M We (Duo2000) is proposed as a new design concept for large nuclear power plant. Duo is being designed to meet economic and safety challenges facing the 21 century green and sustainable energy industry. Duo2000 has two nuclear steam supply systems (NSSS) of the unit nuclear optimizer (Uno) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. Uno is anchored to the optimized power reactor 1000 M We (OPR1000) of the Korea Hydro and Nuclear Power Co., Ltd. The concept of Duo can be extended to any number of PWRs or pressurized heavy water reactors (PHWR s), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the small and medium sized reactors (SMRs) be built as units, the concept of Duo2000 will apply to SMRs as well. With its in-vessel retention as severe accident management strategy, Duo can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for generation III + nuclear systems. The strengths of Duo2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting NSSS. The technology can further be extended to coupling modular reactors as dual, triple, or quadruple units to increase their economics, thus accelerating the commercialization as well as the customization of SMRs. (Author)

  15. Reactor vessel pressure transient protection for pressurized water reactors

    International Nuclear Information System (INIS)

    During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

  16. Atomic data for heavy element impurities in fusion reactors. Summary report of the second research coordination meeting

    International Nuclear Information System (INIS)

    Ten experts on the properties of heavy elements of relevance to fusion energy research participated in the second Research Coordination Meeting (RCM) on Data for Heavy Element Impurities in Fusion Reactors, held at IAEA Headquarters on 26-28 September 2007. Participants summarized their accomplishments with respect to the work plan formulated at the first RCM. This overall work plan was reviewed in detail, achievements were noted, and the plan was subsequently modified to reflect the current state of research. Discussions, conclusions and recommendations of the RCM are briefly described in this report. (author)

  17. 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)

  18. Can light water reactors be proliferation resistant?

    International Nuclear Information System (INIS)

    During the last decade several questions were raised concerning the proliferation issues of Light Water Reactors (LWRs) in comparison with other types of power reactors, particularly Gas Cooled Reactors (GCRs) and Heavy Water Reactors (HWRs). These questions were strongly highlighted when the Agreed Framework between the United States and the DPRK was signed in October 1994 and following the formation of KEDO organization to provide two LWRs to DPRK in replacement of all its GCRs in its nuclear program. One might summarize the main questions into three groups, mainly: 1. Can LWRs produce weapon-grade Plutonium (Pu)? 2. Why is the LWR type considered as a better option with regard to non-proliferation compared to other power reactors - particularly GCR and HWR types? 3. How could LWRs be more resistant to proliferation? This paper summarizes the effort to answer these questions. Included tables present numerical parameters for Pu production capability of the three main reactor types (LWRs, GCRs and HWRs) of a 400 MWe power reactor unit, during normal operation, and during abnormal operation to produce weapon grade Pu. Can LWRs produce weapon-grade Pu? It is seen from the available data that weapon-grade Pu could be produced in LWR fuel, as in the fuel of most other power reactor types, by limiting fuel irradiation to two or three months only. However, such production, though possible, is exceptional. In a recent study 5% of LWRs under IAEA safeguards have spent fuel inventory containing limited amount of high-grade Pu. The equilibrium burnup of discharged fuel is in the order of 33,000 MWD/T. However and due to lower enrichment of initial inventory almost half of that burnup is produced. In normal situations the discharged initial inventory has a Pu grade which is less than weapon grade and is unlikely to be used for weapon production. Why LWR the type is considered as a better option for non-proliferation Referring to tables, one can conclude that LWRs make less Pu

  19. Reconcentration technology of heavy water used as moderator and coolant at Cernavoda Plant

    International Nuclear Information System (INIS)

    One of the basic conditions to ensure the efficient operation of the CANDU reactor plant is maintaining a high isotope purity of heavy water in the moderator and cooling system. It is thus necessary that concentration be 99,80% mol. D2 O for the the moderator, and 99,75% mol. D2 O for the coolant, with water making up the rest. When the nuclear power plant is in service, heavy water in the moderator and cooling circuits may via various ways become mixed with H2 O impurities. Therefore, an isotope purification and, reconcentration of these fluids is necessary. A heavy-water reconcentration technology involving isotope distillation in vacuum in columns filled with phosphorous bronze, was developed by the Institute of Cryogenics and Isotope Separation Rm. Valcea, based on research and experience regarding heavy water concentration in the distillation pilot

  20. Types of Nuclear Reactors

    International Nuclear Information System (INIS)

    The presentation is based on the following areas: Types of Nuclear Reactors, coolant, moderator, neutron spectrum, fuel type, pressurized water reactor (PWR), boiling water reactor (BWR) reactor pressurized heavy water (PHWR), gas-cooled reactor, RBMK , Nuclear Electricity Generation,Challenges in Nuclear Technology Deployment,EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR.

  1. 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.)

  2. 78 FR 46378 - La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact...

    Science.gov (United States)

    2013-07-31

    ... COMMISSION La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact... of Title 10 of the Code of Federal Regulations (10 CFR) for the La Crosse Boiling Water Reactor... modifying or adding EP requirements in Section 50.47, Section 50.54, and Appendix E of 10 CFR part 50 (76...

  3. Progress in the Development of a Heavy-Water Moderated and Cooled Thorium-Uranium-233 Converter

    International Nuclear Information System (INIS)

    On account of its excellent neutron economy the heavy-water reactor is suitable for development as a high-gain converter or as a breeder in the thorium-uranium-233 cycle. In the Federal Republic of Germany work on these lines is being carried out at the Jülich Nuclear Research Centre in co-operation with the Siemens Company. Prominence is being given to a pressure-vessel reactor moderated and cooled by heavy water. Contrary to natural uranium types the thorium reactor will have a quasi-homogeneous core structure and a much lower heavy-water content and therefore will be in appearance more similar to a light-water reactor. With a quasi-homogeneous lattice and the fact that flux peaks are more easily avoided in a D2O core than in an H2O one, the mean specific fuel power (which is decisive for the economic efficiency of a thorium reactor) attains a higher value than in reactors with clustered fuel elements or in light-water reactors. The D2O-thorium reactor will have core power densities in the range of those attained in boiling light-water reactors. This power density is greater than in any other advanced converter. Development work can be based on the well-known technology of water-cooled reactors and on operating experience with the multipurpose reactor (MZFR) at Karlsruhe. The total time and cost required for the development of the heavy-water reactor towards a thorium-uranium-converter or breeder is therefore relatively small. This type will profit in particular from the great progress made in recent years with Zircaloy-canned ceramic fuel rods and with fuel element development. At minimum fuel cycle costs in the range of 1 mill/kWh, including costs for D2O inventory, the specific consumption of 235U for maintaining a reprocessed equilibrium cycle is calculated to be 0.20 g 235U/MWd, including diffusion tails. Calculation methods are shown to be in good agreement with lattice experiments carried out in the Siemens Argonaut Reactor at Garching. On account of the

  4. Boiling water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA

  5. 1000 tones of heavy water produced at ROMAG PROD, Drobeta-Turnu Severin

    International Nuclear Information System (INIS)

    On May 25, 2001 the heavy water plant ROMAG PROD at Drobeta-Turnu Severin recorded the production of the 1000-th tone of nuclear purity heavy water. The heavy water plant ROMAG PROD makes use of a technology based on the results of isotopic deuterium separation research carried out at the Research and Design Institutes of Cluj, Craiova, Pitesti and Ploiesti during 1957-1970 and the separation technology tested at Ramnicu-Valcea pilot plant (at present the Cryogenics and Isotope Separation Institute). The first investments at ROMAG PROD were made in 1979 and on July 17, 1988 was produced the first amount of heavy water at the required parameters for CANDU type nuclear reactors. The period between 1990-1992 was dedicated to the project completion, upgrading the technological facilities and retrofitting the environmental protection and monitoring systems. Production was resumed in 1992. The first 500 t of heavy water required for the Cernavoda NPP first reactor operation were produced by summer 1997. The additional amount of 500 t of heavy water was produced between 1997-2001. ROMAG PROD obtained the ISO 9001/2001 certificate for the quality management system, the ISO 14001/1997 certificate for the environmental management system and the new environmental permit

  6. 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.

  7. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Visa, Maria, E-mail: maria.visa@unitbv.ro; Chelaru, Andreea-Maria, E-mail: andreea.chelaru1@yahoo.com

    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 SiO{sub 2}/Al{sub 2}O{sub 3} 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.

  9. 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.

  10. 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.

  11. 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)

  12. 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.

  13. 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...

  14. 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

  15. Radiation damage of light water reactor materials

    International Nuclear Information System (INIS)

    Reactor materials and irradiation conditions have changed as reactor types advanced from light water reactors to fast breeder reactors and future nuclear fusion reactors, but the physical and chemical processes of irradiation damage at the final stage of service of the structural materials of light water reactors must be studied more in detail. The research on the secular change of LWR materials under irradiation is regarded as most important. Especially in the problems of the embrittlement of pressure vessels, the irradiation damage of in-core structures, and the irradiation growth and creep of zircaloy, it has become more important to continuously evaluate the secular deterioration and assure the soundness, and to elucidate the process of microstructure chanage of the materials under irradiation for forecasting the life. The structure and the materials of light water reactors are explained. The neutron irradiation embrittlement of reactor pressure vessels must be forecast more accurately as the years of operation elapse. The factors affecting the irradiation embrittlement are described. The physical model and the microstructure of the irradiation embittlement are shown. As for in-core structures, intergranular stress corrosion cracking is discussed. As for zirconium alloy, the factors affecting the irradiation creep and growth are explained. (K.I.)

  16. 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

  17. 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.)

  18. 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)

  19. Water injection device for reactor container

    International Nuclear Information System (INIS)

    A pressure vessel incorporating a reactor core is placed and secured on a pedestal in a dry well of a reactor container. A pedestal water injection line is disposed opened at one end in a pedestal cavity passing through the side wall of the pedestal and led at the other end to the outside of the reactor container. A substitution dry well spray line is connected to a spray header disposed at the upper portion of the dry well. When the pressure vessel should be damaged by a molten reactor core and the molten reactor core should drop to the dry well upon occurrence of an accident, the molten reactor core on the floor of the pedestal is cooled by water injection from the pedestal water injection line. At the same time, the elevation of the pressure and the temperature in the reactor container is suppressed by the water injection of the substitution dry well spray line. This can avoid large scaled release of radioactive materials to the environmental circumference. (I.N.)

  20. 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.

  1. Water desalination using different capacity reactors options

    International Nuclear Information System (INIS)

    The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity, cogeneration of potable water production and nuclear electricity is an option to be assessed. In this paper we will perform an economical comparison for cogeneration using a big reactor, the AP1000, and a medium size reactor, the IRIS, both of them are PWR type reactors and will be coupled to the desalination plant using the same method. For this cogeneration case we will assess the best reactor option that can cover both needs using the maximum potable water production for two different desalination methods: Multistage Flash Distillation and Multi-effect Distillation. (authors)

  2. 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)

  3. Single purpose reactor for sea water desalination

    International Nuclear Information System (INIS)

    Some possibilities of a single-purpose reactor for sea water desalination are outlined. Preliminary economic evaluations are also presented and emphasis is given to the prospects of a simplified reactor for sea water desalination. Because no more than 100M3/year are required in one place at one time and given the lack of experience in operating very large desalination plants, it seems that the single-purpose reactor should be small (between 200 mwt to 600 mwt). Two new concepts for desalination plants have been recently developed in Israel: an aluminium horizontal tube multieffect evaporator (AHTME) designed and manufactured by Israel Desalination Engineering Ltd. and a direct contact condensation (DCC) plant, whose 50,000 GPD pilot plant is under construction in the city of Haifa. These two concepts of desalination plants are characterized by economy and operating temperatures below 100 deg C. For the AHTME, the optimum water cost corresponds to a steam temperature of about 70 deg C. A water-cooled reactor can be employed without the need of pressurized vessels or tubes; in addition, cheap construction materials such as aluminium can be used. The advantages of combining a simplified reactor and improved desalination plants and the advantages of the single purpose reactor could bring about a cheaper cost for desalinated water. (author)

  4. The D and D of the Experimental Boiling Water Reactor (EBWR)

    International Nuclear Information System (INIS)

    Argonne National Laboratory has completed the D ampersand D of the Experimental Boiling Water Reactor. The Project consisted of decontaminating and for packaging as radioactive waste the reactor vessel and internals, contaminated piping systems, miscellaneous tanks, pumps, and associated equipment. The D ampersand D work involved dismantling process equipment and associated plumbing, ductwork drain lines, etc., performing size reduction of reactor vessel internals in the fuel pool, packaging and manifesting all radioactive and mixed waste, and performing a thorough survey of the facility after the removal of activated and contaminated material. Non-radioactive waste was disposed of in the ANL-E landfill or recycled. In January 1996 the EBWR facility was formally decommissioned and transferred from EM-40 to EM-30. This paper will discuss the details of this ten year effort

  5. TA-2 water boiler reactor decommissioning (Phase 1)

    International Nuclear Information System (INIS)

    Removal of external structures and underground piping associated with the gaseous effluent (stack) line from the TA-2 Water Boiler Reactor was performed as Phase I of reactor decommissioning. Six concrete structures were dismantled and 435 ft of contaminated underground piping was removed. Extensive soil contamination by 137Cs was encountered around structure TA-2-48 and in a suspected leach field near the stream flowing through Los Alamos Canyon. Efforts to remove all contaminated soil were hampered by infiltrating ground water and heavy rains. Methods, cleanup guidelines, and ALARA decisions used to successfully restore the area are described. The cost of the project was approximately $320K; 970 m3 of low-level solid radioactive waste resulted from the cleanup operations

  6. Neutronic study of a light water core reflected with heavy water

    International Nuclear Information System (INIS)

    For the design of the high flux reactor with extracted beams, we have performed an experimental study on a light water and highly enriched uranium core surrounded by a thick heavy water reflector. This study was made with a critical facility specially designed. The scope of the measurements was double: i) to check the calculation methods then available ii) to evaluate effects not easily calculated owing to the reactor geometry. The main experiments were the following - criticality measurements for various core geometries - measurement of void and temperature coefficients - distribution of neutron density and fission sources - study of kinetic parameters and calibration of control elements - measurements of spectral indices and of flux perturbations due to the presence in the reflector of channels for the neutron beam extraction. The primary analysis of the experimental results gives generally a satisfactory agreement with calculations and shows that in some cases better hypotheses should be taken. (authors)

  7. 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 ...

  8. Requirements for light water reactors

    International Nuclear Information System (INIS)

    The EUR (European Utilities Requirements) is an organization founded in 1991 whose aim was to write down the European specifications and requirements for the future reactors of third generation. EUR gathers most of the nuclear power producers of Europe. The EUR document has been built on the large and varied experience of EUR members and can be used to elaborate invitations to tender for nuclear projects. 4000 requirements only for the nuclear part of the plant are listed, among which we have: -) the probability of core meltdown for a reactor must be less than 10-6 per year, -) the service life of every component that is not replaceable must be 60 years, -) the capacity of the spent fuel pool must be sufficient to store 10-15 years of production without clearing out. The EUR document is both open and complete: every topic has been considered, it does not favor any type of reactor but can ban any technology that is too risky or has an unfavourable feedback experience. The assessment of the conformity with the EUR document of 7 reactor projects (BWR 90/, EPR, EP1000, SWR1000, ABWR, AP1000 and VVER-AES-92) has already be made. (A.C.)

  9. Feasibility study of self sustaining capability on water cooled thorium reactors for different power reactors

    International Nuclear Information System (INIS)

    Thorium fuel cycle can maintain the sustainable system of the reactor for self sustaining system for future sustainable development in the world. Some characteristics of thorium cycle show some advantages in relation to higher breeding capability, higher performance of burn-up and more proliferation resistant. Several investigations was performed to improve the breeding capability which is essential for maintaining the fissile sustainability during reactor operation in thermal reactor such as Shippingport reactor and molten salt breeder reactor (MSBR) project. The preliminary study of breeding capability on water cooled thorium reactor has been investigated for various power output. The iterative calculation system is employed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000. In this calculation, 1238 fission products and 129 heavy nuclides are employed. In the cell calculation, 26 heavy metals and 66 fission products and 1 pseudo FP are employed. The employed nuclear data library was JENDL 3.2. The reactor is fueled by 233U-Th Oxide and it has used the light water coolant as moderator. Some characteristics such as conversion ratio and void reactivity coefficient performances are evaluated for the systems. The moderator to fuel ratio (MFR) values and average burnups are studied for survey parameter. The parametric survey for different power outputs are employed from 10 MWt to 3000 MWt for evaluating the some characteristics of core size and leakage effects to the spectra profile, required enrichment, breeding capability, fissile inventory condition, and void reactivity coefficient. Different power outputs are employed in order to evaluate its effect to the required enrichment for criticality, breeding capability, void reactivity and fissile inventory accumulation. The obtained value of the conversion ratios is evaluated by using the equilibrium atom composition. The conversion ratio is employed based on the

  10. 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

  11. Simulation and optimal operation of a heavy water upgrader plant

    International Nuclear Information System (INIS)

    The paper presents the computer programs elaborated to determine the efficiency of D2 O upgrader plant from Cernavoda NPP and its optimum operating by the right choosing of the feed location of top bottom extraction flow rate for a given reflux. The software is currently used in the operation of the moderator and thermal agent upgrader's systems of the first reactor CANDU in commercial operation in Romania. Since the programs run under Windows operating environment there are many advantages: user friendly graphic interface, easy work with the menus and the dialogue windows, the use of the mouse, the work with different printers and error messages. This software is useful also for who is operating with heavy water final rectification facility using under vacuum distillation in large plants which produce nuclear material. For easy visualization of the results, OPTREC program can display the plant scheme with input and output data (flow rates, concentrations, optimum feed data , etc). (authors)

  12. Use of dried ion exchange resin for heavy water system

    International Nuclear Information System (INIS)

    In order to prevent degradation of D2O in HANARO reflector system due to the moisture in the ion exchange resin, a method using the dried resin is developed. The physical change of dried resin was observed and measured. The performance was tested, and verified. The moisture content in the resin could be reduced to below 1% from its original content of about 55%. The integrated degradation of D2O for 20 years is estimated as 0.23% if the dried resin is used whenever it is replaced. This is much simpler process than the deuteration method which has been used in the other facilities such as heavy water reactors, and the cost of which is almost negligible. Should the dried resin be used for an existing deuteration facility, the generation of degraded D2O will be significantly reduced

  13. 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.

  14. 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)

  15. Safety aspects of pressurised water reactors

    International Nuclear Information System (INIS)

    This submission to the Health and Safety Executive has been prepared by the Institution of Professional Civil Servants (IPCS) as a contribution to the debate on safety aspects associated with Pressurized Water Reactors (PWRs). Although supporting an energy policy which includes the development of nuclear power, assurances are sought on a number of safety issues if it is decided that this should be generated by a PWR-type reactor. These issues are listed. In particular the following are mentioned: the wider publication of design information, the use of elastic-plastic fracture mechanics as the basis for determining pressure vessel integrity, the failure rate of steam generating units, water coolant quality control, greater investigation of two-phase flow accident conditions, the components of the reactor cooling system and training of reactor personnel in the understanding of LOCA effects. (U.K.)

  16. Heavy Water Vodka: flawless purity and award winning design

    OpenAIRE

    Storås, Arill; Sandbu, Kristoffer; Eriksen, Sandra Kristine; Isene, Frode

    2008-01-01

    This thesis presents the findings from an explorative research study of the Chinese vodka market. The research has been conducted on behalf of Heavy Water International, a Norwegian vodka producer. Heavy Water International believes that the vodka market in China has considerable growth potential the next 5-10 years, and wants to enter the Chinese vodka market within 2008. Their first goal is to establish a good distributor-connection and create brand awareness. Heavy Water ...

  17. 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)

  18. 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.

  19. 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

  20. Stability monitoring for boiling water reactors

    Science.gov (United States)

    Cecenas-Falcon, Miguel

    1999-11-01

    A methodology is presented to evaluate the stability properties of Boiling Water Reactors based on a reduced order model, power measurements, and a non-linear estimation technique. For a Boiling Water Reactor, the feedback reactivity imposed by the thermal-hydraulics has an important effect in the system stability, where the dominant contribution to this feedback reactivity is provided by the void reactivity. The feedback reactivity is a function of the operating conditions of the system, and cannot be directly measured. However, power measurements are relatively easy to obtain from the nuclear instrumentation and process computer, and are used in conjunction with a reduced order model to estimate the gain of the thermal-hydraulics feedback using an Extended Kalman Filter. The reduced order model is obtained by estimating the thermal-hydraulic transfer function from the frequency-domain BWR code LAPUR, and the stability properties are evaluated based on the pair of complex conjugate eigenvalues. Because of the recursive nature of the Kalman Filter, an estimate of the decay ratio is generated every sampling time, allowing continuous estimation of the stability parameters. A test platform based on a nuclear-coupled boiling channel is developed to validate the capability of the BWR stability monitoring methodology. The thermal-hydraulics for the boiling channel is modeled and coupled with neutron kinetics to analyze the non-linear dynamics of the closed-loop system. The model uses point kinetics to study core-wide oscillations, and normalized modal kinetics are introduced to study out-of-phase oscillations. The coolant flow dynamics is dominant in the power fluctuations observed by in-core nuclear instrumentation, and additive white noise is added to the solution for the channel flow in the thermal-hydraulic model to generate noisy power time series. The operating conditions of the channel can be modified to accommodate a wide range of stability conditions

  1. Cost effective water treatment program in Heavy Water Plant (Manuguru)

    International Nuclear Information System (INIS)

    Water treatment technology is in a state of continuous evolution. The increasing urgency to conserve water and reduce pollution has in recent years produced an enormous demand for new chemical treatment programs and technologies. Heavy water plant (Manuguru) uses water as raw material (about 3000 m3/hr) and its treatment and management has benefited the plant in a significant way. It is a fact that if the water treatment is not proper, it can result in deposit formation and corrosion of metals, which can finally leads to production losses. Therefore, before selecting treatment program, complying w.r.t. quality requirements, safety and pollution aspects cost effectiveness shall be examined. The areas where significant benefits are derived, are raw water treatment using polyelectrolyte instead of inorganic coagulant (alum), change over of regenerant of cation exchangers from hydrochloric acid to sulfuric acid and in-house development of cooling water treatment formulation. The advantages and cost effectiveness of these treatments are discussed in detail. Further these treatments has helped the plant in achieving zero discharge and indirectly increased cost reduction of final product (heavy water); the dosage of 3 ppm of polyelectrolyte can replace 90 ppm alum at turbidity level of 300 NTU of raw water which has resulted in cost saving of Rs. 15-20 lakhs in a year beside other advantages; the change over of regenerant from HCl to H2SO4 will result in cost saving of at least Rs.1.4 crore a year besides other advantages; the change over to proprietary formulation to in-house formulation in cooling water treatment has resulted in a saving about Rs.11 lakhs a year. To achieve the above objectives in a sustainable way the performance results are being monitored. (author)

  2. 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

  3. 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

  4. 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)

  5. Advancement of light water reactor technology

    International Nuclear Information System (INIS)

    The Japanese technology of light water reactors is based on the technology imported from abroad around 1970, and the experience has been accumulated by the construction, operation and repair of light water reactors as well as the countermeasures to various troubles, moreover, the improvement and standardization of light water reactors have been promoted. As the result, recently the high capacity ratio has been attained, and the LWR technology has firmly taken root in Japan. The Subcommittee for the Advancement of Light Water Reactor Technology of the Advisory Committee for Energy has examined the subjects of technical development and the way the development should be in order to decide the strategy to advance LWR technology, and drawn up the interim report. The change of situation around the LWRs in Japan and the necessity to advance the technology, the target of advancing LWR technology and the subjects of the technical development, the system for the technical development and the securement of fund, and international cooperation are reported. The subjects of development are the pursuit of higher reliability and economic efficiency, the extension of plant life, the improvement of repairability and the reduction of radiation exposure, the improvement of operational capability, the reduction of wastes, the techniques for reactor decommissioning and the diversified location. (Kako, I.)

  6. Thermal or epithermal reactor

    International Nuclear Information System (INIS)

    In a thermal or epithermal heavy-water reactor of the pressure tube design the reactivity is to be increased by different means: replacement of the moderator by additional rods with heavy metal in the core or in the reflector; separation of the moderator (heavy water) from the coolant (light water) by means of shroud tubes. In light-water reactor types neutron losses are to be influenced by using the heavy elements in different configurations. (orig./PW)

  7. The projects for heavy water production of the Argentine National Atomic Energy Commission

    International Nuclear Information System (INIS)

    The bases and scope of the projects for heavy water production that are being currently developed by the Argentine National Atomic Energy Commission (CNEA) are described. As an introduction, the following points are presented: a) the fundamentals of heavy water utilization in a nuclear reactor, with a mention of its properties and uses, b) a review of the physicochemical bases of the principal methods for heavy water production: chemical exchange (monothermal and bithermal processes), distillation and electrolysis, with tables summarizing the fundamental characteristics of the first two ones, and an evaluation of the different production methods from the viewpoint of their application in an industrial scale; and c) a synthetic information, in the form of tables, about the world's heavy water production. The subject of heavy water production in Argentina is treated in the principal section, describing the scope, location, main characteristics and chemical processes corresponding to the projects being developed by CNEA, which currently are the installation of an Industrial Plant in Arroyito (Province of Neuquen), purchased on a turnkey basis and using the NH3/H2 isotopic exchange method; the installation of an Experimental Plant in Atucha (Province of Buenos Aires), for the development of the domestic technology of heavy-water production by the SH2/H2O isotopic exchange method, and the development of the engineering of an industrial plant (''Module 80''), based on the Experimental Plant's technology. (M.E.L.)

  8. 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

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

    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 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

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

    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 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

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

    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 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

  12. 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

  13. 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.

  14. Loss of coolant accident analysis of supercritical water-cooled reactor fuel qualification test loop

    International Nuclear Information System (INIS)

    The supercritical water-cooled reactor fuel qualification test (SCWR-FQT) intends to test a small scale fuel assembly under supercritical water environment in a research reactor. The modified ATHLET code was applied to model the supercritical water-cooled experimental loop containing this fuel assembly and to perform the calculation analysis of the loss of coolant accident induced by the coolant pipe break. The results indicate that the design of existing safety system can practically ensure the effective cooling of the fuel rod experimental section in the accident scenario. The results also show that the modified ATHLET code has good suitability in simulation of supercritical water-cooled system. (authors)

  15. Safety aspects of water chemistry in light water reactors

    International Nuclear Information System (INIS)

    The goals of the water chemistry control programmes are to maximize operational safety and the availability and operating life of primary system components, to maximize fuel integrity, and to control radiation buildup. To achieve these goals an effective corporate policy should be developed and implemented. Essential management responsibilities are: Recognizing of the long-term benefits of avoiding or minimizing: a) system corrosion; b) fuel failure; and c) radiation buildup. The following control or diagnostic parameters are suitable performance indicators: for PWR primary coolant circuits: pH of reactor water (by operating temperature); Concentration of chlorides in reactor water; Hydrogen (or oxygen) in reactor water. For PWR secondary coolant circuits: pH in feedwater; Cation productivity in steam generator blowdown; Iron concentration in feedwater; Oxygen concentration in condensate. And BWR coolant circuits: Conductivity of reactor water; Concentration of chlorides in reactor water; Iron concentration in feedwater; Copper concentration in feedwater. The present document represents a review of the developments in some Member States on how to implement a reasonable water chemistry programme and how to assess its effectiveness through numerical indicators. 12 figs, 20 tabs

  16. A dual pressurized water reactor producing 2000 MWe

    International Nuclear Information System (INIS)

    The Dual Unit Optimizer 2000 MWe (DUO2000) is proposed as a new design concept for large nuclear power plant. DUO is being designed to meet economic and safety challenges facing the 21. century green and sustainable energy industry. DUO2000 has two nuclear steam supply systems (NSSSs) of the Unit Nuclear Optimizer (UNO) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. UNO is anchored to the Optimized Power Reactor 1000 MWe (OPR1000). The concept of DUO can be extended to any number of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the Small and Medium sized Reactors (SMRs) be built as units, the concept of DUO2000 will apply to SMRs as well. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for Generation III+ nuclear systems. Also, the strengths of DUO2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS. Two prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The Coolant Unit Branching Apparatus (CUBA) is proposed

  17. 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

  18. Natural uranium lattice in heavy water

    International Nuclear Information System (INIS)

    A group of Laplacian determinations have been made under critical running conditions in a heavy water pile specially constructed to this end using either complete lattices or samples of lattices employing a two-zone method. The experimental equipment is briefly described: it has been devised to allow rapid modifications of the charge. The methods of measurement employed are also summarily described one operates either by flux charts in the case of lattices which are then used as references, or by progressive replacement of the bars by concentric rings and measurements of the reactivity. In this case, one attempts to obtain the difference between the material laplacian of the central unknown lattice and that of the reference lattice. The method has been specially develop ped to give precision. Results of Laplacian measurements for all these lattice types are presented, allowing the construction of a set of curves as a function of the separation. Various other effects have also been measured: the equivalent reactivity of a mm of water - anisotropy - temperature effect, etc. However in this first attack on the problem, the measurement of a large variety of Laplacian has been carried out, rather than careful measurements in particular cases. It is in this spirit that the interpretation of the results has been made. As a large number of very complex phenomena still escape the possibilities of the calculation, it is considered that a certain number of adjustments are necessary; now these can only give the desired efficiency in forecasting results if they refer to a sufficiently great number of experimental data. It is necessary then to connect the measurements closely on with the other whilst, at the same time, subdividing them according to logically deduced formulae. The principal source of trouble has been that of coherence. The rules governing the calculations employed in the interpretation of the data are given. In the first instance simple formula are used: first of

  19. 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

  20. Coolant mixing in pressurized