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Sample records for enel-8 reactor

  1. ENEL gushes

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    This article reviews the 1995 Annual Report and Accounts for ENEL for 1995. This successful company enjoyed an 11% rise in net profits, with total sales revenues up by 8.2%. Staff numbers were down, but ENEL's staff were paid more overall. (UK)

  2. Enel waiting for the state to act first

    International Nuclear Information System (INIS)

    Anon

    2005-01-01

    Enel will only answer the basic question whether Slovenske elektrarne will complete the reactors 3 and 4 in nuclear power plant in Mochovce next year. Initial analyses made by Enel indicate that the completion of Mochovce would cost over 60 billions Sk (1,54 billions EUR). Since National Property Fund has approved the strategic investment plan two more conditions have to be met before Enel can acquire Slovenske elektrarne. The first condition is that a part of the company assets has to taken over by state owned joint stock company GovCo. This regards reactor V-1 in nuclear power plant Jaslovske Bohunice that should be made ready for closure by end of 2008, damaged reactor A-1, nuclear equipment decommissioning and disposal unit VYZ and hydroelectric power plant in Gabcikovo. The other condition that still has to be fulfilled is passing of a new act on disposal of nuclear power generating equipment. The new act should distribute the fund's resources amongst the individual nuclear energy sources. M. Arcelli (regional manager of Enel for Slovakia) stresses that the money nuclear power station in Mochovce and V-2 earned and will earn during their time in operation will have to remain in the fund to cover their later liquidation i.e. should not be used on liquidation of reactors V-1 or A-1. (authors)

  3. Research and development activity at ENEL for next generation reactors

    International Nuclear Information System (INIS)

    Fornaciari, P.

    1992-01-01

    Italy, a world leader in nuclear electricity production in the mid sixties, became the only large industrialized country which renounced nuclear energy; a fact hardly understandable for a country that, in addition has a worrisome dependence on imported foreign energy. The paper reports the political events that brought Italy to a condition not easy to be sustained in the long run, citing in particular: the so-called 'reconversion' of the Alto Lazio nuclear plant into a fossil fuelled plant, the closure of the Caorso plant after the encouraging outcome of an IAEA inspection, and finally, the contrasting resolutions voted on the matter by the two Parliament chambers. The paper then deals with the issues which need to be solved to allow a nuclear renaissance in Italy. The research program on future reactor designs has been carried out by ENEL (the Italian National Electricity Board) in the framework of a wide international co-operation with the ambitious goal to demonstrate that, even in the case of a severe accident, no population evacuation, nor any long term limitation as to ground utilization is to be planned. On this important issue a large international consensus is growing worldwide. The willingness of ENEL to remain part of the nuclear community has been also proved by its adhesion to the World Association of Nuclear Operators (WANO), created to enhance the operational safety of nuclear plants worldwide

  4. Enel Graduates in Slovakia

    International Nuclear Information System (INIS)

    Galvani, L

    2009-01-01

    Enel constructs two reactors of second generation in Mochovce. Commissioning is expected at the end of 2013, a few months after Italy returns to nuclear. The power plant relies on the support of the public in the country, free way to adopted safety requirements from Brussels. So did inform assigned redactor Laura Galvagni in the edition of the most read Italian economic daily Il Sole 24 Ore on 24 November 2009. (author)

  5. ENEL overall PWR plant models and neutronic integrated computing systems

    International Nuclear Information System (INIS)

    Pedroni, G.; Pollachini, L.; Vimercati, G.; Cori, R.; Pretolani, F.; Spelta, S.

    1987-01-01

    To support the design activity of the Italian nuclear energy program for the construction of pressurized water reactors, the Italian Electricity Board (ENEL) needs to verify the design as a whole (that is, the nuclear steam supply system and balance of plant) both in steady-state operation and in transient. The ENEL has therefore developed two computer models to analyze both operational and incidental transients. The models, named STRIP and SFINCS, perform the analysis of the nuclear as well as the conventional part of the plant (the control system being properly taken into account). The STRIP model has been developed by means of the French (Electricite de France) modular code SICLE, while SFINCS is based on the Italian (ENEL) modular code LEGO. STRIP validation was performed with respect to Fessenheim French power plant experimental data. Two significant transients were chosen: load step and total load rejection. SFINCS validation was performed with respect to Saint-Laurent French power plant experimental data and also by comparing the SFINCS-STRIP responses

  6. Pioneering corporate responsibility in Italy: the Enel experience

    Energy Technology Data Exchange (ETDEWEB)

    Comin, Gianluca

    2007-07-01

    Enel, the largest Italian utility for production, distribution of electricity and the second largest for natural gas distribution has started its Corporate Responsibility practice in 2002 and ever since has gained a leading role within the Italian sustainability scene. By rapidly integrating Corporate Responsibility in its strategic planning and in its reporting procedures, Enel also could develop a state of the art sustainability plan, a wide array of key performance indicators to monitor it and an outstanding knowledge in Corporate Responsibility mainstreaming. The acceleration imposed by Enel's management in implementing Corporate Responsibility as a basic integrator into its corporate procedures also allowed the company to be acknowledged as one of the best practices in the world. Along the implementation of Corporate Responsibility Enel also developed traditional and intranet based training programs that impacted on 4,500 middle managers and 48,000 employees. International sustainability indices recognised the company's effort including it in their listings as of 2003. Active in stakeholder dialogue, Enel is also implementing a way to gather fresh and direct opinions from an esteemed universe of 60,000 people with an unprecedented web based polling tool derived from multidimensional scaling techniques aimed at questioning appreciation on 25 topics of mutual concern. (auth)

  7. Evolution of Italian national electric power system: Relationship between ENEL and local operators

    International Nuclear Information System (INIS)

    Rossi, G.P.

    1993-01-01

    With the aim of facilitating the option to privatize the Italian national electricity system known as ENEL (Italian National Electricity Board), new laws have been passed which have modified the legal framework pertaining to the management of ENEL's activities and have given ENEL a new status of state concessionaire. The modifications to the existing laws, however, were made in such a way as not to jeopardize the present integral nature of ENEL's activities relative to power generation and distribution. This paper discusses the impact of the new laws on the management of municipal and industrial power plants run by public administrations and private industrial firms licensed to operate under existing and future agreements made with ENEL

  8. ENEL's DSM actions in the international scenario

    International Nuclear Information System (INIS)

    Oliva, P.; Russo, S.

    1997-01-01

    In the framework of national policies which, since the end of the 70s, have had the target of making consumers rationally use the natural sources, in particular energy, and in the pursuit of environmental goals, the Governments and the utilities have driven several DSM (Demand-Side Management) actions, in many countries. The activities implemented by ENEL have included the analysis of the different components of the energy chain, from the primary sources to the final uses in industry, agriculture, residential and commercial buildings. This is in order to improve the energy efficiency by new electrotechnologies and more effective appliances. ENEL has carried out commercial initiatives by specific contractual agreements based on the marginal costs of the electric service. The paper explains ENEL's strategies and the actions implemented by the utility on DMS; the results in terms of energy efficiency, electric demand dynamics and of load curve changing are also explained. Finally the paper explores the DSM perspectives in the Italian future scenario, where energy companies will act in a more competitive market. (author)

  9. Nitrogen oxide control at power plants of the ENEL company (Italy)

    Energy Technology Data Exchange (ETDEWEB)

    Kotler, V.R. (Vserossiiskii Teplotekhnicheskii Institut (Russian Federation))

    1993-03-01

    Analyzes experiences of the ENEL electricity company in Italy in controlling pollutant emission from fossil-fuel power plants. In 1990, the company produced 87% of the country's electricity. Until the year 2000, ENEL plans to increase coal use for power generation by 23.5% and install 9,300 MW of new coal-fired power plant capacity. New European and Italian emission standards require ENEL to reduce NO[sub x] emissions by 30% from 1986 to 1998. NO[sub x] emission values from various fuel-oil and pulverized-coal fired steam generators operated by the company are given. Modifications to existing combustion technologies and equipment installed to lower NO[sub x] content in flue gases at various ENEL power plants are considered. The most promising coal combustion technologies and ongoing research programs are pointed out. 4 refs.

  10. Enel seeks strategic re-entry into the photovoltaic market. Cooperations promise economic efficiency; Enel strebt strategischen Wiedereinstieg in die Photovoltaik an. Kooperationen versprechen Wirtschaftlichkeit

    Energy Technology Data Exchange (ETDEWEB)

    Berger, W. [Orizzonti, Verona (Italy)

    2008-12-01

    Major Italian utility Enel has embarked again in activities in the field of photovoltaics (pv) after years of standstill caused by the company's privatisation in 1999. Now attractive feed-in tariffs, much untapped potential and strong interest of Italians in pv have opened the door for the development of new business in this unfolding mass market. More recently Enel has also strengthened ties with pv-cell and module production industry. (orig.)

  11. SE and ENEL - One Year Together One Year after Privatization

    International Nuclear Information System (INIS)

    Anon

    2007-01-01

    In Bratislava, one year ago, exactly in April 28, the privatization deal of a 66% stake in Slovenske elektrarne (SE) was finally closed. The Slovak government and the Italian company ENEL signed final documents and ENEL thus definitely entered into SE. In May 2006, Miroslav Pikus, a former CEO, was replaced by Marco Arcelli, who, as a regional manager for Slovakia, had been preparing the entry of ENEL into SE since March 2005. Not only did he sometimes cook for us, in his 'kitchen' but also there were prepared 160 investment projects, e.g the improvement of nuclear safety and resistance of the Bohunice power plant, electricity sale through the so-called 'virtual power plant', ensuring a non-attended operation in hydro power plants, reconstruction and modernization of the Orava and Dobsina hydro power plants. (author)

  12. ENEL's experience in the management of irradiated fuel. Old and new problems encountered by nuclear station operators

    International Nuclear Information System (INIS)

    Ariemma, A.; Cuzzaniti, M.; Zaffiro, B.; Bertini, A.

    1983-01-01

    The experience acquired in recent years in the management of spent fuel discharged from ENEL's reactors has revealed a substantial change in the attitude of nuclear station operators in connection with the end of the fuel cycle downstream from the reactor (back-end). While in the past, after President Carter's outline of US policy in April 1977, the utilities had to face great difficulty in reprocessing their fuel owing to inadequate capacity, today the same problem is regarded as a matter of cost-benefit analysis from an industrial standpoint and of appropriate planning for the utilization of the recovered fissile materials. Since the present technology allows spent fuel storage (dry or underwater) to be planned for rather long periods and plutonium utilization requires a very stringent schedule, the present trend is to ensure medium-term storage of spent fuel and to seek a greater flexibility in the final reprocessing stages so as to render plutonium availability consistent with the programmes for its utilization. As a consequence, the solution to the problems posed by high-activity waste disposal is being delayed, thus allowing an exhaustive and detailed analysis of all the possible solutions to be made. The paper describes a number of solutions to the problems ENEL has encountered in the fuel cycle back-end. (author)

  13. What Enel and Rusko will not find in the privatization documents

    International Nuclear Information System (INIS)

    Janoska, J.

    2004-01-01

    The lawyers of Enel will be spending the pre-Christmas period studying the transaction documents relating to the privatization of a 66% stake in Slovenske elektrarne, a.s. (Slovak Electric). At the beginning of the year, the Italian company Enel is set to sign the document that Cabinet approved in December 2004. Neither the draft privatization documents, nor the version approved by Parliament are publicly available. The Ministry promised to publish them after Enel signed the contracts. But Cabinet-level discussions prior to the approval of the documents uncovered problematic parts of the agreements related to the relationship between Enel, Slovak Electric and the state. Lawyers are examining the changes that were implemented based on recommendations of Cabinet members during the period after the postponement of the signing of the contract. Minister Pavol Rusko does not expect that Enel will reject the approved conditions, as that would mean the end of negotiations. But despite this, TREND asked whether he had an alternative plan prepared for such an eventuality. The Minister repeated that he had no doubt that the sale would be completed successfully and as of June Enel would become the new majority shareholder of Slovak Electric. He believes the 'the contract, which was very thoroughly prepared, will be very acceptable for both the selling and purchasing parties'. However, both the lawyers of Enel and the Minister should be aware of new facts they cannot be found in the privatisation documentation. And this is good news for both parties. This week Slovak Electric completed the process of financial restructuring and fulfilled the 'stabilization scenario'. Debt repayment has been postponed, debts will be repayable under standard conditions and will reflect the current capacity of the loan market. State guarantees have been decreased to a minimum and Slovak Electric will be able to request a new rating. More good news, mainly for P. Rusko, although he has not spoken

  14. ENEL Slovenske elektrarne. Annual report 2006

    International Nuclear Information System (INIS)

    2007-01-01

    A brief account of activities carried out by the ENEL Slovenske elektrarne, a.s. (SE) in 2006 year is presented. These activities are reported under the following headings: (1) Evaluation Report and Foreword of the Chairman of the Board of Directors; (2) SE, a.s., Board of Directors in 2006; (3) SE, a.s., Supervisory Board in 2006; (4) The year in brief; (5) SE, a.s., Organisation Structure; (6) Company stakes; (7) Company strategy; (8) Electricity and heat generation; (9) Trade in electricity, support services and heat; (10) Investments; (11) Nuclear safety; (12) Occupational health and safety and fi re protection; (13) Environment; (14) Human resources; (15) Report on Financial Statements and Performance of the Company; (16) Auditor's report; (17) Abbreviations

  15. More electric power for cleaner air: ENEL air pollution abatement program

    International Nuclear Information System (INIS)

    Landini, P.; Segreto, F.

    1991-01-01

    ENEL (Italian National Electricity Board) is conducting an intensive anti-pollution campaign that is expected to lead to overall reductions by about 80% in polluting emissions by the year 2000. Existing plants are being equipped with low NOx burners and electrostatic precipitators; coal units are receiving flue gas desulfurization systems. New poly-fuel plants, to be located far from residential zones, are to use low sulfur fuels or natural gas, and employ 250 meter high smoke-stacks. ENEL's anti-pollution campaign also involves a series of co-ordinated research studies directed towards the development of new sulfur and nitrogen oxides emission control technologies and innovative fluidized-bed and coal-water fuel slurry combustion systems. ENEL will also participate with the Ministry of Agriculture and Forests in environmental protection programs. These will involve the setting-up of acid rain monitoring and sampling networks and atmospheric circulation mapping and modelling

  16. Rahvastikuregister peaks olema keskne / Enel Pungas ; interv. Alo Lõhmus

    Index Scriptorium Estoniae

    Pungas, Enel

    2003-01-01

    Intervjuu siseministeeriumi rahvastikutoimingute osakonna juhataja Enel Pungasega riigiasutuste andmebaaside kooskõlastamisest, rahvastikuregistrist ja e-riigist. Vt. samas lühiintervjuud rahvastikuregistri töötajaga: Eesti e-riik ei ühti iseendaga

  17. Italian cogeneration legislation (Laws 9 ampersand 10): ENEL's (Italy) point of view

    International Nuclear Information System (INIS)

    Pello', P.M.

    1991-01-01

    Referring to recently legislated Italian normatives governing on-site electric power generation by private industry, in particular, to the clauses which regard rate structure in the case of auto-producers ceding power to the national grid controlled by ENEL (the Italian National Electricity Board), this paper determines auto-production economic feasibility limits, based on cogeneration plant annual power production, ceded power quality, type of fuel, and overall operating costs. Some consideration is then given to the long term implications that this legislation, encouraging on-site production on a wide scale, has on ENEL's strategic planning and natural gas marketing in Italy

  18. An example o French - Italian cooperation in the area of energy: ENEL in France

    International Nuclear Information System (INIS)

    Cremieux, M.

    2010-01-01

    The electricity market liberalization also allows to renew European industrial cooperation. As a symbol of such cooperation, Enel, the first both Italian and Spanish operator, develops an integrated position in France where industrial cooperation with Electricite de France in the EPR project goes with a business activity competing with the historical operator. Such industrial cooperation allows Enel to benefit by experience in all available technologies and is of crucial importance in light of the Italian government's choice to develop nuclear energy in Italy. Those industrial agreements, beneficial both for consumers and for the European development, cannot develop unless a number of economic properties are complied with. (author)

  19. David and Goliath : the procedures for selling ENEL distribution lines to city-owned companies

    International Nuclear Information System (INIS)

    Onofri, L.

    2005-01-01

    This paper analyzed the economic impacts of the provisions under Article 9 of the Bersani Decree, relating to the sale of the ENEL distribution networks to city-owned companies, with particular reference to the ENEL-AEM Milano case. The Bersani Decree states that the sale of networks must be made according to normal market rules, but also rules that ENEL is obliged to sell its own distribution network. Through the use of analytical instruments inspired by the contract theory, this study highlighted several contradictions in the decree. It was noted that the deferment to private parties bargaining to perform the sale was contradictory, as it was counterbalanced and blocked by the legal compulsion to sell. The Bersani Decree influences the bargaining power of the parties by handing it over to the city-owned company, who then make as low an evaluation of the contract as possible. This has prompted several companies to seek arbitration. If the arbitrators are unable to arrive at a satisfactory price, the contract will not be completed. The refusal to accept a proposed price may be seen as an attempt to negotiate a different purchase price. In addition, magistrates undertake the bargaining power that would be be granted to both parties to respect the Bersani Decree, in that the magistrate has the power to order one or both parties to conclude the contract. The independent setting of the sale is compromised and does not necessarily reflect the evaluations of parties involved. Appeals to appraisers, hearings and research mean that the costs to both parties, and society, increases. It was suggested that legislators should allow the market to work independently, without mandating the terms of private companies' transactions, and without additional procedural phases. A market-oriented formulation could avoid useless extensions and costs in the sale of ENEL-owned distribution networks to city-owned companies. 17 refs., 4 tabs., 12 figs

  20. Socio-economic and environmental sustainability estimate of ENEL project for the S. Maria in Brasimone basin area; Valutazione della sostenibilita' socio-economica e ambientale del progetto ENEL per l'area del bacino di S. Maria del Brasimone

    Energy Technology Data Exchange (ETDEWEB)

    Padovani, L; Beone, F; Carrabba, P [ENEA, Centro Ricerche Casaccia, Rome (Italy). Dipt. Ambiente; Lorenzelli, E; Cialani, C

    1999-07-01

    During preparatory activities of National Conference on Energy and Environment (CNEA), which took place in Rome in November 1998, the ENEA Biological Diversity Group and ENEL signed an agreement to lead a socio-economic and environmental sustainability analysis of a tourism valorisation project about an ENEL vacant, marginal, fragile or abandoned site. The activities had also the aim to be an effective contribution to individuation of future guidelines on sustainable tourism which the Convention on Biological Diversity (CBD) will discuss and issue, if possible. This work has the purpose of proposing a territorial dialogue example between many parties, which would start a local development that was sustainable, respecting local traditions. Actions proposed by ENEA are: 1. integration of a project based on environmental valorisation to tourist purpose, adding elements of respect and sustainable use of local biological diversity. 2. Analysis, through a series of medium/long term statistic model, both of the interventions proposed by ENEL and of the possible integrations suggested by ENEA. 3. Collaboration in searching for institutional funds (UE), for the achievement of the demonstrative project on sustainable tourism as regards of local biological diversity. The analysis conducts by ENEA allowed the following conclusions: the cost-benefit analysis, for the period taken into account, highlight as the project could have a positive result only if it is possible to rely on Regional or European funds. The possibility to obtain these types of funds in the fields of tourism is proportional to the consideration given into the project to the environment, to the rural development and the diversification of the income in the rural areas. Moreover, the authors suggests that the ENEL project should be wider, inclusive of more interventions, as i.e. the restoration of old building to turn into agritourism, recreational centres and hotels: the improvement of particular local

  1. Socio-economic and environmental sustainability estimate of ENEL project for the S. Maria in Brasimone basin area; Valutazione della sostenibilita' socio-economica e ambientale del progetto ENEL per l'area del bacino di S. Maria del Brasimone

    Energy Technology Data Exchange (ETDEWEB)

    Padovani, L; Beone, F.; Carrabba, P. [ENEA, Centro Ricerche Casaccia, Rome (Italy). Dipt. Ambiente; Lorenzelli, E.; Cialani, C.

    1999-07-01

    During preparatory activities of National Conference on Energy and Environment (CNEA), which took place in Rome in November 1998, the ENEA Biological Diversity Group and ENEL signed an agreement to lead a socio-economic and environmental sustainability analysis of a tourism valorisation project about an ENEL vacant, marginal, fragile or abandoned site. The activities had also the aim to be an effective contribution to individuation of future guidelines on sustainable tourism which the Convention on Biological Diversity (CBD) will discuss and issue, if possible. This work has the purpose of proposing a territorial dialogue example between many parties, which would start a local development that was sustainable, respecting local traditions. Actions proposed by ENEA are: 1. integration of a project based on environmental valorisation to tourist purpose, adding elements of respect and sustainable use of local biological diversity. 2. Analysis, through a series of medium/long term statistic model, both of the interventions proposed by ENEL and of the possible integrations suggested by ENEA. 3. Collaboration in searching for institutional funds (UE), for the achievement of the demonstrative project on sustainable tourism as regards of local biological diversity. The analysis conducts by ENEA allowed the following conclusions: the cost-benefit analysis, for the period taken into account, highlight as the project could have a positive result only if it is possible to rely on Regional or European funds. The possibility to obtain these types of funds in the fields of tourism is proportional to the consideration given into the project to the environment, to the rural development and the diversification of the income in the rural areas. Moreover, the authors suggests that the ENEL project should be wider, inclusive of more interventions, as i.e. the restoration of old building to turn into agritourism, recreational centres and hotels: the improvement of particular local

  2. ENEL (Italy) 3.3 MW photovoltaic power plant at Serre

    International Nuclear Information System (INIS)

    Iliceto, A.; Previ, A.; Corsi, S.; Belcastro, G.N.; Vigotti, R.

    1993-01-01

    This paper describes the key design and performance features of a demonstration 3.3 MW photovoltaic power plant which is being built by ENEL (the Italian National Electricity Board) to determine the technical-economic feasibility of such installations and to augment national engineering know-how and construction capability in this field. The plant is to be made up of 60,000 modules, each equipped with sets of 36 or 72 10x10 cm polycrystalline silicon cells. The paper also provides sketches of module support systems, main electrical layouts and computerized control systems, as well as, a breakdown of project costs

  3. Upgrading of operation and maintenance. Lessons learned by ENEL- Nuclear Plant Management

    International Nuclear Information System (INIS)

    Bolla, Giuseppe; Panini, Giulio

    1999-01-01

    A presentation and general information on ENEL are given. The Italian Authority for Electric Energy (Ente Nazionale per l'Energia Electrica - ENEL), entirely owned by the Italian Ministry of Treasury, is the third largest utility in the world after EDF of France and TEPCO of Japan. The following figures are featuring its size: - Net maximum capacity = 52.463 GW; - Power generation = 182,500 GW h; - Customer = 28.2 Million; - Hydroelectric power plants = 651 (16.327 MW); - Thermoelectric power units = 206 (35,666 MW); - Geothermal power units = 36 (470 MW). The ENEL core business consists of generation, transmission, distribution and sale throughout Italy of electric power produced from any source as well as the importation and exportation of electric power and the co-ordination of operational and planning activities in order to ensure adequate availability throughout Italian territory of electric power at minimum cost. Before the decision taken for political reason by Italian Government in 1988 to terminate nuclear power plants operation and construction, 4 NPPs have been successfully operated. The whole production of the four plants has been about 90 billion kWh. The management experience for these plants has been good, both for plant availability and safety. No operating events occurred arising any risk for workers, population and environment. The Chernobyl power plant accident induced Italy to promote the 'nuclear referendum' in November 1987. Following the referendum the nuclear power production was interrupted. After 1988 the following nuclear activities has been carried out: - plant safe conservation, decommissioning planning, cost evaluation and safe enclosure preparation; - investigation for waste treatment methods; - participation in requirement definition and design activity for next NPP generation. A specific experience has been accumulated in the decommissioning of Italian NPPs, due to their anticipated shutdown, in particular in the field of

  4. A review of fast reactor activities in Italy

    International Nuclear Information System (INIS)

    Pierantoni, F.; Tavoni, R.

    1990-01-01

    Italy's total primary energy demand increased in 1989 by 2.8% continuing the trend initiated in 1984. Net electricity import increased by 9% reaching the total of 34 TWh, equivalent to the electrical production of 5 power stations in base load operation. As known, after the Chernobyl accident, a moratorium of at least 5 years was imposed on the construction of new nuclear reactors in Italy. The PEN (National Energy Plan), already approved by the government and submitted to parliament for discussion, calls for a programme of research and industrial development of reactors with inherent and passive safety features. In this framework during 1989, a collaboration of ENEL, Ansaldo and ENEA with some designers of innovative reactors, mainly water-cooled, has taken place in many fields. The Italian participation and contribution in international groups has continued with the aim of maintaining and updating the know-how on the most relevant themes of research. Contacts are underway in the area of advanced reactors with several countries such as the USA, the FRG, the UK and France. With France an agreement for cooperation in the area of future nuclear reactors and advanced technologies for the upgrading of industries was signed in April 1990. (author). 1 fig., 2 tabs

  5. A review of the Italian fast reactor programme, March 1979

    Energy Technology Data Exchange (ETDEWEB)

    Pierantoni, F

    1979-07-01

    Year 1978 in Italy was marked by a standstill in the nuclear energy field. The decisions previously made for the installation of eight 1000 MWe LWR-type reactors could not be acted upon because of the opposition of local authorities and lack of Government power. The construction site at Montalto di Castro (two BWR reactors) was ope ned with difficulty, whereas the decision to install a plant in Mo use equipped with two PWR reactor was postponed. The new presidents of ENEL and CNEN were appointed in January this year and the appointments of the new Boards of Directors are underway. With regard to CNEN, many political bodies are in agreement on an institutional change which would widen field of activity to include new energy sources, solar energy in particular. This will open a big problem: if CNEN will be no more a 'nuclear body, it could be necessary to transfer all the activities connected to the Regulatory Commission to another separate body to be instituted. In this context, the fast reactor programme has continued to develop under the directives of CIPE, and has concentrated its effort on the following three objectives: the PEC-Reactor, the Creys-Malville Power Plant and research and development, and industrial promotion. These objectives are being pursued with the participation of CNEN, ENEL and Italian industry. CNEN has the role of committing and operating the PEC reactor; it is also charged to perform part of the R and D Italian-French programme and to promote industrial development. ENEL participates in the NERSA Company, owner of the Creys-Malville Plant. Italian industry, with its activities of architect-engineering, designing and manufacturing will participate in the construction of the PEC and of the Italian part (33%) of the Creys-Malville Plant. During the last months of 1978 a consortium (COREY) was set up by CNEN and NIRA which has the purpose of integrating and ensuring the smooth running of Italian efforts in the field of long-term research and

  6. A review of the Italian fast reactor programme, March 1979

    International Nuclear Information System (INIS)

    Pierantoni, F.

    1979-01-01

    Year 1978 in Italy was marked by a standstill in the nuclear energy field. The decisions previously made for the installation of eight 1000 MWe LWR-type reactors could not be acted upon because of the opposition of local authorities and lack of Government power. The construction site at Montalto di Castro (two BWR reactors) was ope ned with difficulty, whereas the decision to install a plant in Mo use equipped with two PWR reactor was postponed. The new presidents of ENEL and CNEN were appointed in January this year and the appointments of the new Boards of Directors are underway. With regard to CNEN, many political bodies are in agreement on an institutional change which would widen field of activity to include new energy sources, solar energy in particular. This will open a big problem: if CNEN will be no more a 'nuclear body, it could be necessary to transfer all the activities connected to the Regulatory Commission to another separate body to be instituted. In this context, the fast reactor programme has continued to develop under the directives of CIPE, and has concentrated its effort on the following three objectives: the PEC-Reactor, the Creys-Malville Power Plant and research and development, and industrial promotion. These objectives are being pursued with the participation of CNEN, ENEL and Italian industry. CNEN has the role of committing and operating the PEC reactor; it is also charged to perform part of the R and D Italian-French programme and to promote industrial development. ENEL participates in the NERSA Company, owner of the Creys-Malville Plant. Italian industry, with its activities of architect-engineering, designing and manufacturing will participate in the construction of the PEC and of the Italian part (33%) of the Creys-Malville Plant. During the last months of 1978 a consortium (COREY) was set up by CNEN and NIRA which has the purpose of integrating and ensuring the smooth running of Italian efforts in the field of long-term research and

  7. Valorización de Enel Generación Piura S.A.

    OpenAIRE

    Graham, Matthew; Quintanilla Pacheco, Mario; Sulca Gómez, Lucio

    2017-01-01

    Enel Generación Piura S.A. tiene como actividad principal la generación de energía eléctrica con un negocio secundario de venta de gas natural. La generación de energía eléctrica es la primera actividad en la cadena de producción y suministro de energía y por su naturaleza pertenece a un mercado competitivo y altamente regulado. En la industria de la generación es fundamental la disponibilidad y continuidad de las fuentes de materia prima, así como las diversas tecnologías que se encuentran e...

  8. Enel expects privatization to be completed

    International Nuclear Information System (INIS)

    Anon

    2006-01-01

    The collapse of the coalition will probably mean that the sale of state-owned stakes in two distribution companies and all the heating plants will not be completed and nor will privatisation at Zapadoslovenska energetika. Most important for the energy sector is uncertainty regarding the major privatization project of this government - finding a new owner of 66% of Slovenske elektrarne shares. The state expects to receive about 32 billion Slovak crowns (842.105 million EUR) from Enel for this stake. 'The cabinet has approved the privatization under certain conditions and those have to be met now,' answered the leader of the KDH deputy's club, Pavol Minarik, to the question whether the cabinet would still be able to muster sufficient votes in parliament to pass the Act on the national nuclear account. All other so-called delaying conditions are in the competence of the cabinet or the Ministry of Economy of the Slovak Republic and the opposition, which opposes privatization, cannot influence them. The act will address the financing of debts related to the liquidation of the V1 and A1 nuclear power plants. (authors)

  9. ENEL high and medium voltage electrical substations for power supply to urban centers

    International Nuclear Information System (INIS)

    Bargigia, A.; Boatto, C.; Di Mario, A.; Fava, N.; Sciarra, S.; Speziali, R.

    1991-12-01

    Modular high and medium voltage gas insulated electrical substations are being used by ENEL (Italian Electricity Board) to meet the specific needs of urban centers with special design and sizing constraints of a historical/architectural nature. This paper illustrates the key design, construction operation and performance characteristics of these standardized units. The descriptions include brief notes on equipment-transformer interconnection, interchangeability and environmental compatibility. Performance test procedures and results of actual reliability and certification tests on some substations are reported

  10. Past and present role of fast breeder reactors in Italy

    International Nuclear Information System (INIS)

    Castelli, G.; Cicognani, G.; Ghilardotti, G.; Musso, B.

    1978-01-01

    The paper describes the programme that is under development in Italy for fast breeder reactors. The Italian engagement in the construction of the Creys-Malville plant is discussed as well as the work at present under way in R and D activities and on the PEC reactor. The roles of the different organizations involved in the fast breeder reactor programme are also considered with particular attention to the activities carried out by AGIP NUCLEARE, CNEN, ENEL, NIRA and by the manufacturing companies. An overall picture of the different agreements between European countries is also given with reference to the construction of the Creys-Malville plant and to research and industrial development activities. (author)

  11. ENEL and application of Italian November 14, 1990 Decree CIP 34/3

    International Nuclear Information System (INIS)

    Vazio, A.

    1991-01-01

    As part of an overall energy conservation campaign, a concrete step, in the form of a more favourable electricity rate structure for auto-producers, is being taken by the Italian Government to encourage medium sized industries to adopt cogeneration systems to meet their heat and power requirements. Within this context, this paper gives an idea of how ENEL, the Italian National Electricity Board, administers contracts involving the ceding of on-site generated power to the national utility's grid. Reference is made to the specific provisions incorporated in the CIP (Italian Inter-ministerial Commission on Prices) Provision No. 34, dated November 14, 1990

  12. Pu utilization in fast-breeder and in light-water reactors in Italy

    International Nuclear Information System (INIS)

    Mangiagalli, D.; Cicognani, F.; Pistella, F.; Testa, G.; Villani, A.; Ariemma, A.; Castelli, G.F.; Linari, A.; Paoletti Gualandi, M.; Musso, B.

    1977-01-01

    The paper illustrates the most important activities carried out in Italy for the development of fast breeder reactors and its fuel as well as for plutonium recycle in light water reactors. The Italian strategy is based, on one hand, on the short-term commercialization of fast breeder reactors, and on the other, on the adoption of the technology of the Phenix prototype whose further development will be ensured by the joint Italian and French efforts as insured by the important agreements signed by CNEN, NIRA (Nucleare Italiana Reattori Avanzati) and Italian manufacturing industries with CEA and the main French industries. The paper also includes the main results of the ENEL Demonstration Program on Pu prototypes introduced in the Garigliano BWR in 1968 and 1970, and of the destructive and non-destructive analyses on said fuel, as well as of the analyses carried out by CNEN on prototypical fuel fabricated by CNEN and irradiated in various reactors. Furthermore, the paper deals with design and licensing aspects of the 46 Pu-island assembly reload introduced in the Garigliano reactors in 1975 and of a batch of 8 all Pu assemblies loaded in the Trino Vercellese PWR in 1976. Subsequently, the experimental activities planned for the near future both on high burn-up prototypes and on industrial fuel after one cycle of operation are examined [fr

  13. Reactor Power Meter type SG-8

    Energy Technology Data Exchange (ETDEWEB)

    Glowacki, S W

    1981-01-01

    The report describes the principle and electronic circuits of the Reactor Power Meter type SG-8. The gamma radiation caused by the activity of the reactor first cooling circuit affectes the ionization chamber being the detector of the instrument. The output detector signal direct current is converted into the frequency of electric pulses by means of the current-to-frequency converter. The output converter frequency is measured by the digital frequency meter: the number of measured digits in time unit is proportional to the reactor power.

  14. The contribution of nuclear energy to meeting Italy's electric power needs

    International Nuclear Information System (INIS)

    Angelini, A.M.

    1977-01-01

    ENEL's policy in the development of nuclear energy is based on a number of factors which differentiate Italy's situation from that of most industrialized countries. Among these are the scarcity of natural resources, including energy sources, available in Italy. An extensive recourse to nuclear energy is the best way to achieve that diversification of energy supplies vital to the Italian economy, which, because of the very lack of natural resources, is predominantly based on processing. ENEL's nuclear plan calls for the construction of the four 1000-MW units already on order and of 16 additional units of the same size, divided into two 8-unit blocks, to be decided on respectively in the very near future. The necessary flexibility of the plan for nuclear units to go into service by 1986 is ensured by the subdivision into blocks, with provision for the postponement of the second 8-unit block should the country's economic development require a revision in electric power forecasts. The paper considers in particular the integration between the nuclear plan and ENEL's extensive plan for pumped-storage hydroelectric power plants and the related technical and economic advantages, also on an international scale. The review of Italy's nuclear plan concludes by stressing two essential problems: financing and the availability of sites for nuclear power plants. Finally, the paper supplies some information on the nuclear research, development and demonstration activities connected with the long-term plans; they are centred on the CNEN-ENEL project for the 40,000-kW CIRENE prototype and above all on the EdF-ENEL-RWE joint project for the construction of two over-1000-MW demonstration plants equipped with sodium-cooled breeder reactors. With such reactors it would be possible to achieve a virtually complete energy independence for the nuclear source. (author)

  15. The story of the European fast reactor cooperation

    International Nuclear Information System (INIS)

    Marth, W.

    1993-12-01

    This report is a condensed history of European cooperation in the large breeder power plants with powers in excess of 1000 MWe. The beginning, in 1973, was marked by the so-called Utilities' Convention signed by EdF, RWE, and ENEL on the construction of Superphenix and SNR 2. In 1977, cooperation began among the reactor vendors and R and D organizations in France, Germany and Italy as well as Belgium and the Netherlands. After the British had joined in 1984, planning for the European Fast Reactor, EFR, was started in 1988. The conceptual design phase of the 1500 MWe breeder power plant covered a period of five years and was concluded with an economic assessment and a technical safety analysis of EFR in 1983. A number of ongoing studies are being conducted within a specific EFR program. (orig.) [de

  16. Comparison between MAAP and ECART predictions of radionuclide transport throughout a French standard PWR reactor coolant system

    International Nuclear Information System (INIS)

    Hervouet, C.; Ranval, W.; Parozzi, F.; Eusebi, M.

    1996-04-01

    In the framework of a collaboration agreement between EDF and ENEL, the MAAP (Modular Accident Analysis Program) and ECART (ENEL Code for Analysis of radionuclide Transport) predictions about the fission product retention inside the reactor cooling system of a French PWR 1300 MW during a small Loss of Coolant Accident were compared. The volatile fission products CsI, CsOH, TeO 2 and the structural materials, all of them released early by the core, are more retained in MAAP than in ECART. On the other hand, the non-volatile fission products, released later, are more retained in ECART than in MAAP, because MAAP does not take into account diffusion-phoresis: in fact, this deposition phenomenon is very significant when the molten core vaporizes the water of the vessel lower plenum. Centrifugal deposition in bends, that can be modeled only with ECART, slightly increases the whole retention in the circuit if it is accounted for. (authors). 18 refs., figs., tabs

  17. Enhanced distribution automation system based on new communication technology at ENEL

    International Nuclear Information System (INIS)

    Comellini, E.; Gargiuli, R.; Gelli, G.; Tonon, R.; Mirandola, P.

    1991-01-01

    On the basis of extensive research aimed at the assessment of the feasibility of a cost effective two-way telecommunications system, and of the experience gained during the eighties in the field of remote control of the primary distribution network where new digital techniques were introduced, and in the field of metering apparatus, where about 7,000 HV and MV customers were equipped with Ferraris meters associated with electronic devices for the application of multirate tariffs, ENEL (Italian National Electricity Board) has designed a new distribution automation system aimed at: remote control of the MV distribution network, and MV and LV customer meter service automation. This report describes the key choices that determined the architecture of the new system and the most important features of its main components, in view of: an improvement of energy usage efficiency, better service to the customers, as well as, increased simplicity and transparency in customer relationships

  18. Enhanced distribution automation system based on new communications technology at ENEL

    Energy Technology Data Exchange (ETDEWEB)

    Comellini, E; Gargiuli, R; Gelli, G; Tonon, R; Mirandola, P

    1992-12-31

    On the basis of extensive research aimed at the assessment of the feasibility of a cost effective two-way telecommunications system, and of the experience gained during the eighties in the field of remote control of the primary distribution network where new digital techniques were introduced, and in the field of metering apparatus, where about 7,000 HV and MV customers were equipped with Ferraris meters associated with electronic devices for the application of multirate tariffs, ENEL (Italian National Electricity Board) has designed a new distribution automation system aimed at: remote control of the MV distribution network, and MV and LV customer meter service automation. This report describes the key choices that determined the architecture of the new system and the most important features of its main components, in view of: an improvement of energy usage efficiency, better service to the customers, as well as, increased simplicity and transparency in customer relationships.

  19. PV plants for Alpine huts: Installation and operating experience at seven ENEL plants

    International Nuclear Information System (INIS)

    Belli, G.; Iliceto, A.; Previ, A.

    1988-01-01

    The problem of supplying electric power to isolated users far from the electricity distribution grid is one of general interest. Such consumers are nowadays generally supplied with electricity produced by small diesel generator-sets, and only recently have photovoltaic arrays and wind-turbines offered an alternative to the internal combustion engine. ENEL, as a State-owned electricity utility, is interested in the development of this particular application of renewable energy sources. Enlarging a low-voltage distribution network to connect consumers whose power requirements are extremely low (about 1000 kWh/year) may, in certain conditions, be uneconomical, both for the utility, which has to absorb most of the expense involved in construction and maintenance, and for the consumer himself. The paper reports the design criteria, the tests and the problems encountered in electrifying seven alpine huts belonging to CAI (Italian Alpine Club)

  20. Quality of electric service in utility distribution networks under electromagnetic compatibility principles. [ENEL

    Energy Technology Data Exchange (ETDEWEB)

    Chizzolini, P.; Lagostena, L.; Mirra, C.; Sani, G. (ENEL, Rome Milan (Italy))

    1989-03-01

    The development of electromagnetic compatibility criteria, being worked out in international standardization activities, requires the establishment of the characteristics of public utility distribution networks as a reference ambient. This is necessary for gauging the immunity levels towards users and for defining the disturbance emission limits. Therefore, it is a new way to look at the quality of electric service. Consequently, it is necessary to check and specify, in an homogeneous manner, the phenomena that affect electric service. Use must be made of experimental tests and of the collection and elaboration of operation data. In addition to testing techniques, this paper describes the checking procedures for the quality of electric service as they are implemented in the information system developed by ENEL (Italian Electricity Board) for distribution activities. The first reference data obtained from the national and international fields about voltage shape and supply continuity are also indicated.

  1. Overview of standards subcommittee 8, fissionable materials outside reactors

    International Nuclear Information System (INIS)

    McLaughlin, T.P.

    1996-01-01

    The American Nuclear Society's Standards Subcommittee 8, titled open-quotes Fissionable Materials Outside Reactors,close quotes has worked for the past 35 yr to prepare and promote standards on nuclear criticality safety for the handling, processing, storing, and transportation of fissionable materials outside reactors. The reader is referred to the Transactions of the American Nuclear Society, Vols. 39 (1981) and 64 (1991), for previous papers associated with ANS-8 poster sessions. In addition to discussions on the then-current standards, the reader will find articles on working group efforts that never materialized into standards, such as proposed 8.13, open-quotes Use of the Solid-Angle Method in Nuclear Criticality Safety,close quotes and on applications and critiques of current standards. The paper by McLendon in Vol. 39 is particularly interesting as an overview of the early history of ANS-8 and its standards

  2. Advanced diagnostics and predictive maintenance to improve availability and reliability of ENEL plants

    Energy Technology Data Exchange (ETDEWEB)

    Cenci, V.; Ghironi, M.; Guidi, L.; Lauro, M.; Pestonesi, D. [ENEL (Italy). Generation and Energy Management Division

    2007-07-01

    This paper reviews the ENEL Generation and Energy Management strategy for diagnostics and predictive maintenance of power plants and provides a comprehensive description of effective applications and systems. Exploiting the most advanced information and communication technologies makes it possible to capture weak and hidden signals and powerful processing can be used to discover forewarning symptoms and identify anomalies both in the process and, above all, inside the devices. The following systems and applications are presented together with results and impact on plant profitability: expert system for the diagnostics of plant main machinery; advanced diagnostics of 'intelligent' fieldbus devices such as on/off valve motor-driven actuators, control-valve positioners and pneumatic actuators, transmitters; control loop and control valve diagnostics in order to investigate valve friction with an estimation of the residual time to failure; multisensorial diagnostics for coal transport and storage systems aimed at preventing firing and structural damages; and wireless sensor networks for the diagnostics of medium and small size components. 4 refs., 14 figs., 1 tab.

  3. The contribution of nuclear energy to the meeting of Italy's electric power requirements

    International Nuclear Information System (INIS)

    Angelini, A.M.

    1977-01-01

    The paper discusses ENEL's policy in the development of nuclear energy, which assigns to this source the role of meeting almost all of Italy's additional future power requirements. This is a position taken some time ago and already outlined on the occasion of the Geneva Conference of 1971. The policy is based on a number of factors, reviewed in the paper, which differentiate, at least in quantitative terms, Italy's situation from that of most industrialized countries. Among these factors stand out the scarcity of natural resources, including energy sources available in Italy. An extensive recourse to nuclear energy is the best way to achieve that diversification of energy supplies vital to the Italian economy which, because of the very lack of natural resources, is predominantly based on processing. While the Caorso Nuclear Power Plant (840 MW) is due to go into service soon, ENEL's nuclear plan, recently approved by the Government, calls for the construction of the four 1,000-MW units already on order and of 16 additional units, of the same size, divided into two 8-units blocks, to be decided on respectively in the very near future and before the end of 1977. The necessary flexibility of the plan concerning the nuclear units that will go into service by the 1986 is ensured by the subdivision into blocks, with provision for the postponement of the second, 8-units block in the country's economy development requires a revision in electric power forecasts. The paper then considers in particular the integration between the nuclear plan and ENEL's extensive plan for pumped-storage hydro-electric power plants and the related technical and economic advantages which also extend to an international scope. The paper concludes the review of Italy's nuclear plan by stressing two essential problems: financing and the availability of sites for nuclear power plants. Upon their timely and satisfactory solution depends the actual construction of the plants by the scheduled dates

  4. A model of Alto Lazio boiling water reactor using the LEGO code balance of plant simulation

    International Nuclear Information System (INIS)

    Spelta, S.; Garbossa, G.B.

    1989-01-01

    An extensive effort has been made at the Italian National Electricity Board (ENEL) to construct and validate a LEGO model capable of simulating the operational transients of the Alto Lazio Nuclear Station, a two twin-units site with BWR/6 class reactors, rated at 2894 MWt and with Mark III containment. LEGO is a modular package developed at the Research and development Department of the Italian National Electricity Board (CRA-ENEL) for computer aided modeling of fossil-fired and nuclear steam power plants. In this paper a system analysis model capable of describing steady-state and transient performance of the Balance of Plant (BOP) of the Alto Lazio Power Station is presented. This is one of two companion papers devoted to the description of the overall plant model including both the Nuclear Steam Supply System (NSSS) and the BOP. In the paper, after a brief summary of the main LEGO characteristics, a description of the BOP lay-out is presented. The overall model, which has been set-up, including control systems and automation, is very detailed and consists of almost 2000 differential or algebraic equations. After a brief description of the mathematical model, two significant transients obtained using the overall model are presented and discussed

  5. Comparison between MAAP and ECART predictions of radionuclide transport throughout a French standard PWR reactor coolant system; Transport des radionucleides dans le circuit primaire d`un REP: comparaison des codes MAAP et ECART

    Energy Technology Data Exchange (ETDEWEB)

    Hervouet, C.; Ranval, W. [Electricite de France (EDF), 92 - Clamart (France); Parozzi, F.; Eusebi, M. [Ente Nazionale per l`Energia Elettrica, Rome (Italy)

    1996-04-01

    In the framework of a collaboration agreement between EDF and ENEL, the MAAP (Modular Accident Analysis Program) and ECART (ENEL Code for Analysis of radionuclide Transport) predictions about the fission product retention inside the reactor cooling system of a French PWR 1300 MW during a small Loss of Coolant Accident were compared. The volatile fission products CsI, CsOH, TeO{sub 2} and the structural materials, all of them released early by the core, are more retained in MAAP than in ECART. On the other hand, the non-volatile fission products, released later, are more retained in ECART than in MAAP, because MAAP does not take into account diffusion-phoresis: in fact, this deposition phenomenon is very significant when the molten core vaporizes the water of the vessel lower plenum. Centrifugal deposition in bends, that can be modeled only with ECART, slightly increases the whole retention in the circuit if it is accounted for. (authors). 18 refs., figs., tabs.

  6. The qualification of U3O8 as research reactor fuel

    International Nuclear Information System (INIS)

    Krull, W.

    1983-01-01

    This report summarizes the today knowledge of the qualification status of U 3 O 8 as low enriched ( 3 O 8 is so far qualified to start testing of ten (10) fuel elements with an U-density of 3.1 g U/cc in the FRG-2 research reactor. (orig.) [de

  7. 'Thin walled' concept and a new top lid applied to the Scandinavian PCRV for a boiling water reactor

    International Nuclear Information System (INIS)

    Scotto, F.L.

    1975-01-01

    This research is carried out in the frame of an agreement between AB ATOMENERGI of Sweden and ENEL (Ente Nazionale per l'Energia Elettrica) of Italy, for an exchange of information in the field of PCPV for BWR, and takes as a reference the Scandinavian solution as far as the thermal insulation system and the geometry are concerned, proposing new solutions for the prestressed concrete structure (namely the Author's concept of thin walls and a new concept of top lid). The proposed top lid sealing system solution is in line with the one adopted for the conventional steel pressure vessel enclosures; furthermore the prestressed concrete lid is restricted to the prestressed concrete structure to form a continuous contrete structure, in line with th PCPV conventional solutions for gas reactors. The paper describes in detail the selected design philosophy that is slightly different from the one defined by the Scandinavian project. In fact, as far as the design limits are concerned, it refers mainly to steel pressure vessel philosophy and, as to the concrete behaviour, to the design philosophy proposed by the author for the PCPV 'thin walled' structures for gas-cooled power reactors. Rheological, mathematical and physical models had been suitably devised in order to check the reliability of the proposed assumption. This paper therefore, will also give a brief description of said tools and the main results acquired at the time of the conference, and technical and economical considerations made to support the interest of the research, showing the relevant cut down of the costs. The comparative reference steel pressure vessel belongs to Mark III ENEL VI and VIII BW plant to which design and construction the author gives his contribution

  8. A Particular Experience: How a Nuclear Expert Became an Antinuke

    International Nuclear Information System (INIS)

    Ferrari Ruffino, Giorgio

    2017-01-01

    Between 1968 and 1987, I was in charge of control activities on nuclear fuel design and fabrication of all nuclear power plants of ENEL. At the same time, I attended to several cooperation agreements signed between ENEL and Euratom focused on Post Irradiated fuel Examination (PIE), which was the largest PIE program realized in Europe on power reactors. After the oil shock of 1973, my attitude toward nuclear power changed because the role of nuclear energy was not longer the one outlined in the program Atoms for Peace “to provide abundant electrical energy in the power-starved areas of the world”. On the contrary, its role become to strengthen the position of big energy companies and of the countries that controlled this new technology. So I became an antinuke. Throughout this period, I continued to carry out my control activities on ENEL nuclear fuel, until 1987 when, after the Chernobyl disaster, I decided for a moral objection, as I wrote in an official letter to ENEL, making my decision public.

  9. Design and build of fieldbus-based control systems for ENEL thermoelectric plants; Sistemi di controllo con bus di campo. Applicazioni in impianti termoelettrici

    Energy Technology Data Exchange (ETDEWEB)

    Mazzola, G [Automazione e Robotica, Cesi (Italy); Prandoni, W [Business Unit Processi per la Generazione, Milan (Italy)

    2000-06-01

    The intention to start first plant-wide control applications based on fieldbus in ENEL (Electric Power Production Company) combined cycle plants demands to define appropriate requirements for design, planning, implementation, technical management and maintenance of control systems. One of the first needs is to express, in the bid specifications for Dc san other plant components, those requirements necessary on one side to guarantee the same levels of performance and reliability of traditional systems, and on the other side to ensure the improvement of the operations of installation and commissioning and of management of normal and abnormal plant behaviour, which are among the main benefits expected from this new technology. The definition of such requirements appears to be complex due to the variety and lack of a homogenous approach of the different market solutions, and also because a real experience of systems of considerable size is still missing. The memory introduces the results and problems met in the requirement specification phase, and the activity of definition of a pilot control system for demineralized water production of Fusina power plant. [Italian] In vista delle prime applicazioni negli impianti ENEL a ciclo combinato della tecnologia a bus di campo si presenta la necessita' di esprimere, nelle specifiche di acquisizione del Dcs e dei componenti, i requisiti necessari da un lato a garantire i medesimi livelli di prestazione e sicurezza dei sistemi tradizionali, dall'altro ad assicurare il miglioramento delle operazioni di installazione e delle strategie di conduzione e manutenzione dell'impianto, che sono come i ritorni piu' promettenti di questa nuova tecnologia. L'articolo presenta le principali problematiche emerse e l'attivita' di definizione di un'esperienza pilota su un impianto in esercizio nella centrale termoelettrica di Fusina.

  10. Design and build of fieldbus-based control systems for ENEL thermoelectric plants; Sistemi di controllo con bus di campo. Applicazioni in impianti termoelettrici

    Energy Technology Data Exchange (ETDEWEB)

    Mazzola, G. [Automazione e Robotica, Cesi (Italy); Prandoni, W. [Business Unit Processi per la Generazione, Milan (Italy)

    2000-06-01

    The intention to start first plant-wide control applications based on fieldbus in ENEL (Electric Power Production Company) combined cycle plants demands to define appropriate requirements for design, planning, implementation, technical management and maintenance of control systems. One of the first needs is to express, in the bid specifications for Dc san other plant components, those requirements necessary on one side to guarantee the same levels of performance and reliability of traditional systems, and on the other side to ensure the improvement of the operations of installation and commissioning and of management of normal and abnormal plant behaviour, which are among the main benefits expected from this new technology. The definition of such requirements appears to be complex due to the variety and lack of a homogenous approach of the different market solutions, and also because a real experience of systems of considerable size is still missing. The memory introduces the results and problems met in the requirement specification phase, and the activity of definition of a pilot control system for demineralized water production of Fusina power plant. [Italian] In vista delle prime applicazioni negli impianti ENEL a ciclo combinato della tecnologia a bus di campo si presenta la necessita' di esprimere, nelle specifiche di acquisizione del Dcs e dei componenti, i requisiti necessari da un lato a garantire i medesimi livelli di prestazione e sicurezza dei sistemi tradizionali, dall'altro ad assicurare il miglioramento delle operazioni di installazione e delle strategie di conduzione e manutenzione dell'impianto, che sono come i ritorni piu' promettenti di questa nuova tecnologia. L'articolo presenta le principali problematiche emerse e l'attivita' di definizione di un'esperienza pilota su un impianto in esercizio nella centrale termoelettrica di Fusina.

  11. Review of Past Reactor Activities in Italy, April 1975

    International Nuclear Information System (INIS)

    Pierantoni, P.

    1975-01-01

    Since the last meeting of the International Working Group on Fast Reactors the natural vocation of the italian program to a cooperation, illustrated on many occasions, has found factual confirmation. As well-known, a broad partnership with France has been entered into. Four agreements have been concluded, towards May-June 1974 - between C.E.A. and C.N.E.N. for the ulterior development of the French filière and the construction and use of PEC; - between C.E.A. and NIRA for the license on the system; - between C.E.A. and AGIP-NUCLEARE for the fabrication of fuel for fast reactors; - between TECHNICATOME and NIRA for the preparation of the offer and the later construction of Super-Phenix. Those agreements are in the full logic of the previous one passed between the producers E.d.F., ENEL and R.W.E. for the construction of the first two commercial-size plants, extrapolated from Phenix and S.N.R. 300. In particular the NERSA society, acting as client for Super-Phenix, has been founded by those same producers and will receive the offer now prepared by a joint team. The new orientation of the programme and the validity of the previous agreements for its implementation have been confirmed at governmental level by the Interministerial Economic Planning Committee (CIPE), which has also approved the third CNEN's five years Plan

  12. Part II. Chapter VII. Lack of knowledge has generated misconceptions about WWER in Western Europe

    International Nuclear Information System (INIS)

    Bucha, T.

    2008-01-01

    In this interview Mr. Giancarlo Aquilanti (Project Director Mochovce NPP, Unit-3 and Unit-4, the Slovenske elektrarne, a. s., a member of ENEL) evaluates reactor safety in Mochovce NPP. Perspectives of development of nuclear energy in the Slovak Republic and Italia are discussed.

  13. 8th International School of Fusion Reactor Technology "Ettore Majorana"

    CERN Document Server

    Leotta, G G; Muon-catalyzed fusion and fusion with polarized nuclei

    1988-01-01

    The International School of Fusion Reactor Technology started its courses 15 years ago and since then has mantained a biennial pace. Generally, each course has developed the subject which was announced in advance at the closing of the previous course. The subject to which the present proceedings refer was chosen in violation of that rule so as to satisfy the recent and diffuse interest in cold fusion among the main European laboratories involved in controlled thermonuclear research (CTR). In the second half of 1986 we started to prepare a workshop aimed at assessing the state of the art and possibly of the perspectives of muon- catalyzed fusion. Research in this field has recently produced exciting experimental results open to important practical applications. We thought it worthwhile to consider also the beneficial effects and problems of the polarization ofthe nuclei in both cold and thermonuclear fusion. In preparing the 8th Course on Fusion Reactor Technology, it was necessary to abandon the tradi...

  14. Chance for young nuclear professionals in Slovenske Elektrarne, member of Enel Group

    International Nuclear Information System (INIS)

    Zlatnansky, Jozef

    2014-01-01

    Main areas and opportunities for young nuclear professionals: → Further increase of safety of NPPs in operation and → Construction of two new NPP units; → Safety and efficiency of NPP decommissioning and radwaste treatment; → CENTA and R&D; → Prototype of gas cooled fast-breader reactor - ALLEGRO. Gas-cooled fast reactor (GFR) is one of the Generation IV reactor concepts and represents one of the three European candidate for fast reactor types. Conclusions: • Human resources are key assets for nuclear organization; • Knowledge of nuclear technology be managed as a resource; • Support of young nuclear professionals is our commitment

  15. Evaluation of the problems associated with ENEL's irradiated fuel management

    International Nuclear Information System (INIS)

    Sani, L.; Belelli, U.

    1978-01-01

    The present difficulties in the reprocessing field, wich will continue for a period of time not easily definable, impose on the nuclear utilities the research of adequate solutions to solve the problem associated with storage of the fuels discharged by the nuclear power stations in operation. In this context, the paper examines the technical aspects of the various possible alternatives and on the basis of the Italian nuclear program, possible storage strategies are discussed in the assumption of the development of an adequate oxide-fuel reprocessing capacity. The problems related to storage in pools at the station or away from reactor are described in the light of the recent experience made in Italy with the adaptation of an old pool reactor out of service, that is, the Avogadro reactor in Saluggia, to store the fuels of the stations now in operation. Finally some economic aspects are considered in order to ascertain their impact on the cost of energy produced. (author)

  16. Nuclear Reactor RA Safety Report, Vol. 8, Auxiliary system

    International Nuclear Information System (INIS)

    1986-11-01

    This volume describes RA reactor auxiliary systems, as follows: special ventilation system, special drainage system, hot cells, systems for internal transport. Ventilation system is considered as part of the reactor safety and protection system. Its role is eliminate possible radioactive particles dispersion in the environment. Special drainage system includes pipes and reservoirs with the safety role, meaning absorption or storage of possible radioactive waste water from the reactor building. Hot cells existing in the RA reactor building are designed for production of sealed radioactive sources, including packaging and transport [sr

  17. Ten-yearly report on operations at the Garigliano nuclear power station

    International Nuclear Information System (INIS)

    1977-01-01

    This document is the final report on operation of the Garigliano Nuclear Power Station as required under the participation contract between Enel and Euratom and refers to the first ten years of commercial operation (1 May 1964-31 December 1973) of the power station. In the decade in question the Garigliano Power Station has assumed an important position in the world spectrum of nuclear energy since it was the first thermal-power reactor in the world to have irradiated considerable quantities of plutonium as a fuel in its own core for the production of energy. An experimental programme on this was started in 1966 with theoretical study of plutonium recycling and was followed by the charging of the Garigliano reactor with some mixed oxide elements (PuO 2 -UO 2 ) in 1968 and 1970. The excellent performance of these prototype elements, which among other things were examined in detail at the end of each radiation cycle, prompted Enel to decide in favour of the use of entire batches of recycled fuel elements of the PuO 2 -UO 2 type in the reactor from 1975 onwards

  18. A model of Altio Lazio boiling water reactor using the LEGO code nuclear steam supply system simulation

    International Nuclear Information System (INIS)

    Garbossa, G.B.; Spelta, S.; Cori, R.; Mosca, R.; Cento, P.

    1989-01-01

    An extensive effort has been made at the Italian National Electricity Board (ENEL) to construct and validate a LEGO model capable of simulating the operational transients of the Alto Lazio Nuclear Station, a two-twin units site with BWR/6 class reactors, rated at 2894 MWt and with Mark III containment. The desired end-product of this effort is an overall plant model consisting of the Nuclear Steam Supply System model, described in this paper, and the Balance of Plant model, capable of simulating the transient response of Alto Lazio Station. The models utilize the in-house developed LEGO code, which is a modular package oriented to power plant modeling and suitable to perform transient analyses to assist during power plant design, control system design and operating procedure verification. The ability of the NSSS model to predict correctly the plant response is demonstrated through comparison with results calculated by the vendor, using REDY code, and by an in-house RETRAN-02 model

  19. Operational characteristics analysis of a 8 mH class HTS DC reactor for an LCC type HVDC system

    International Nuclear Information System (INIS)

    Kim, S. K.; Go, B. S.; Dinh, M. C.; Park, M.; Yu, I. K.; Kim, J. H.

    2015-01-01

    Many kinds of high temperature superconducting (HTS) devices are being developed due to its several advantages. In particular, the advantages of HTS devices are maximized under the DC condition. A line commutated converter (LCC) type high voltage direct current (HVDC) transmission system requires large capacity of DC reactors to protect the converters from faults. However, conventional DC reactor made of copper causes a lot of electrical losses. Thus, it is being attempted to apply the HTS DC reactor to an HVDC transmission system. The authors have developed a 8 mH class HTS DC reactor and a model-sized LCC type HVDC system. The HTS DC reactor was operated to analyze its operational characteristics in connection with the HVDC system. The voltage at both ends of the HTS DC reactor was measured to investigate the stability of the reactor. The voltages and currents at the AC and DC side of the system were measured to confirm the influence of the HTS DC reactor on the system. Two 5 mH copper DC reactors were connected to the HVDC system and investigated to compare the operational characteristics. In this paper, the operational characteristics of the HVDC system with the HTS DC reactor according to firing angle are described. The voltage and current characteristics of the system according to the types of DC reactors and harmonic characteristics are analyzed. Through the results, the applicability of an HTS DC reactor in an HVDC system is confirmed

  20. Operational characteristics analysis of a 8 mH class HTS DC reactor for an LCC type HVDC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. K.; Go, B. S.; Dinh, M. C.; Park, M.; Yu, I. K. [Changwon National University, Changwon (Korea, Republic of); Kim, J. H. [Daejeon University, Daejeon (Korea, Republic of)

    2015-03-15

    Many kinds of high temperature superconducting (HTS) devices are being developed due to its several advantages. In particular, the advantages of HTS devices are maximized under the DC condition. A line commutated converter (LCC) type high voltage direct current (HVDC) transmission system requires large capacity of DC reactors to protect the converters from faults. However, conventional DC reactor made of copper causes a lot of electrical losses. Thus, it is being attempted to apply the HTS DC reactor to an HVDC transmission system. The authors have developed a 8 mH class HTS DC reactor and a model-sized LCC type HVDC system. The HTS DC reactor was operated to analyze its operational characteristics in connection with the HVDC system. The voltage at both ends of the HTS DC reactor was measured to investigate the stability of the reactor. The voltages and currents at the AC and DC side of the system were measured to confirm the influence of the HTS DC reactor on the system. Two 5 mH copper DC reactors were connected to the HVDC system and investigated to compare the operational characteristics. In this paper, the operational characteristics of the HVDC system with the HTS DC reactor according to firing angle are described. The voltage and current characteristics of the system according to the types of DC reactors and harmonic characteristics are analyzed. Through the results, the applicability of an HTS DC reactor in an HVDC system is confirmed.

  1. Multigroup P8 - elastic scattering matrices of main reactor elements

    International Nuclear Information System (INIS)

    Garg, S.B.; Shukla, V.K.

    1979-01-01

    To study the effect of anisotropic scattering phenomenon on shielding and neutronics of nuclear reactors multigroup P8-elastic scattering matrices have been generated for H, D, He, 6 Li, 7 Li, 10 B, C, N, O, Na, Cr, Fe, Ni, 233 U, 235 U, 238 U, 239 Pu, 240 Pu, 241 Pu and 242 Pu using their angular distribution, Legendre coefficient and elastic scattering cross-section data from the basic ENDF/B library. Two computer codes HSCAT and TRANS have been developed to complete this task for BESM-6 and CDC-3600 computers. These scattering matrices can be directly used as input to the transport theory codes ANISN and DOT. (auth.)

  2. Operating reactors licensing actions summary. Volume 5, No. 8

    International Nuclear Information System (INIS)

    1985-10-01

    This summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

  3. Proceedings of the 8. Brazilian Meeting on Reactor Physics and Thermal Hydraulics

    International Nuclear Information System (INIS)

    1991-01-01

    Some papers about pressurized light water reactors, fast reactors, accident analysis, transients, research reactors, nuclear data collection, thermal hydraulics, reactor monitoring, neutronics are presented. (E.G.)

  4. Feasibility study of the vectorization of nuclear codes used at the ENEL Thermal and Nuclear Research Centre

    International Nuclear Information System (INIS)

    Di Pasquantonio, F.

    1987-01-01

    The purpose of this report is that of analyzing tha problems connected with the vectorization and/or multitasking of several computer codes utilized in the ENEL-DSR Centro Ricerca Termica e Nucleare. After some general remarks on vector computers the analysis is focused on some topic relating to vectorization and multitasking of programs written for scalar computers. The priority for vectorization and/or multitasking has been given at the following codes: 1) DOT 4.2 (radiation transport and shielding); 2) QUANDRY (accidental and operating transients in LWR cores); 3) MORSE and MCNP (Monte Carlo codes for radiation transport and shielding); 4) RELAP (accidental and operating transients in LWR plants); 5) TRAP-MELT and NAUA (evaluation of source term). The principal results of the study are the following: 1) For the DOT 4.2 code it is convenients to improve the vectorized version DOT IV/C developed by Swanson introducing the parallel S.O.R. iterative method; 2) For the code QUANDRY it is proposed to introduce the three dimensional red-black mesh point ordering named ''diagonal method''; 3) To implements, on the CRAY X/MP 48, the multitasked version of the code MCNP developed by the Los Alamos National Lboratory; 4) To implements, on the CRAY X/MP 12, the vectorized and optimized version of the codes RELAP5/MOD1-MOD2 developed by J.R.C. EUROATOM-Ispra; 5) For the codes TRAP-MELT and NAUA the insertion of the vectorized routines LSODP na LSODPK for dominanting stiff cases

  5. Effect of Utilization of Silicide Fuel with the Density 4.8 gU/cc on the Kinetic Parameters of RSG-GAS Reactor

    International Nuclear Information System (INIS)

    Setiyanto; Sembiring, Tagor M.; Pinem, Surian

    2007-01-01

    Presently, the RSG-GAS reactor using silicide fuel element of 2.96 gU/cc. For increasing reactor operation time, its planning to change to higher density fuel. The kinetic calculation of silicide core with density 4.8 gU/cc has been carried out, since it has an influence on the reactor operation safety. The calculated kinetic parameters are the effective delayed neutron fraction, the delayed neutron decay constant, prompt neutron lifetime and feedback reactivity coefficient very important for reactor operation safety. the calculation is performed in 2-dimensional neutron diffusion-perturbation method using modified Batan-2DIFF code. The calculation showed that the effective delayed neutron fraction is 7. 03256x10 -03 , total delay neutron time constant is 7.85820x10 -02 s -1 and the prompt neutron lifetime is 55.4900 μs. The result of prompt neutron lifetime smaller 10 % compare with silicide fuel of 4.8 gU/cc. The calculated results showed that all of the feedback reactivity coefficient silicide core 4.8 gU/cc is negative. Totally, the feedback reactivity coefficient of silicide fuel of 4.8 gU/cc is 10% less than that of silicide fuel of 2.96 gU/cc. The results shown that kinetic parameters result decrease compared with the silicide core with density 2.96 gU/cc, but no significant influence in the RSG-GAS reactor operation. (author)

  6. A neutronic feasibility study for LEU conversion of the IR-8 research reactor

    International Nuclear Information System (INIS)

    Deen, J.R.; Hanan, N.A.; Matos, J.E.; Egorenkov, P.M.; Nasonov, V.A.

    1998-01-01

    Equilibrium fuel cycle comparisons for the IR-8 research reactor were made for HEU (90%), HEU (36%), and LEU (19.75%) fuel assembly (FA) designs using three dimensional multi-group diffusion theory models benchmarked to detailed Monte Carlo models of the reactor. Comparisons were made of changes in reactivity, cycle length, average 235 U discharge burnup, thermal neutron flux, and control rod worths for the 90% and 36% enriched IRT-3M fuel assembly and the 19.75% enriched IRT-4M fuel assembly with the same fuel management strategy. The results of these comparisons showed that a uranium density of 3.5 g/cm 3 in the fuel meat would be required in the LEU IRT-4M fuel assembly to match the cycle length of the HEU (90%) IRT-3M FA and an LEU density of 3.7 g/cm 3 is needed to match the cycle length of the HEU (36%) IRT-3M FA. (author)

  7. Comparison of the parameters of the IR-8 reactor with different fuel assembly designs with LEU fuel

    International Nuclear Information System (INIS)

    Vatulin, A.; Stetsky, Y.; Dobrikova, I.

    1999-01-01

    The estimation of neutron-physical, heat and hydraulic parameters of the IR-8 research reactor with low enriched uranium (LEU) fuel was performed. Two fuel assembly (FA) designs were reviewed: IRT-4M with the tubular type fuel elements and IRT-MR with the rod type fuel elements. UO 2 -Al dispersion 19.75% enrichment fuel is used in both cases. The results of the calculations were compared with main parameters of the reactor, using the current IRT-3M FA with 90% high enriched uranium (HEU) fuel. The results of these comparisons showed that during the LEU conversion of the reactor the cycle length, excess reactivity and peak power of the IRT-MR type FA are higher than for the IRT-3M type FA and IRT-4M type FA. (author)

  8. Study of the U3O8-Al thermite reaction and strength of reactor fuel tubes

    International Nuclear Information System (INIS)

    Peacock, H.B.

    1984-01-01

    Research and test reactors are presently operated with aluminum-clad fuel elements containing highly enriched uranium-aluminum alloy cores. To lower the enrichment and still maintain reactivity, the uranium content of the fuel element will need to be higher than currently achievable with alloy fuels. This will necessitate conversion to other forms such as U 3 O 8 -aluminum cermets. Above the aluminum melting point, U 3 O 8 and aluminum undergo an exothermic thermite reaction and cermet fuel cores tend to keep their original shape. Both factors could affect the course and consequences of a reactor accident, and therefore prompted an investigation of the behavior of cermet fuels at elevated temperatures. Tests were carried out using pellets and extruded tube sections with 53 wt % U 3 O 8 in aluminum. This content corresponds to a theoretical uranium density of 1.9 g/cc. Results indicate that the thermite reaction occurs at about 900 0 C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U 3 O 8 -aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 660 0 C. In tension, sections failed at about the aluminum melting point. In compression with 2 psi average axial stress, failure occurred at 917 0 C, while 7 psi average axial stress produced failure at 669 0 C. (author)

  9. Study of the U3O8-Al thermite reaction and strength of reactor fuel tubes

    International Nuclear Information System (INIS)

    Peacock, H.B.

    1983-01-01

    Research and test reactors are presently operated with aluminum-clad fuel elements containing highly enriched uranium-aluminum alloy cores. To lower the enrichment and still maintain reactivity, the uranium content of the fuel element will need to be higher than currently achievable with alloy fuels. This will necessitate conversion to other forms such as U 3 O 8 -aluminum cermets. Above the aluminum melting point, U 3 O 8 and aluminum undergo an exothermic thermite reaction and cermet fuel cores tend to keep their original shape. Both factors could affect the course and consequences of a reactor accident, and prompted an investigation of the behavior of cermet fuels at elevated temperatures. Tests were carried out using pellets and extruded tube-sections with 53 wt % U 3 O 8 in aluminum. This content corresponds to a theoretical uranium density of 1.9 g/cc. Results indicate that the thermite reaction occurs at about 900 0 C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U 3 O 8 -aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 660 0 C. In tension, sections failed at about the aluminum melting point. In compression with 2-psi average axial stress, failure occurred at 917 0 C, while 7 psi average axial stress produced failure at 669 0 C

  10. Recent developments: Industry briefs

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    Recent nuclear industry briefs are presented. These briefs include: Swiss voters reject phaseout; EdF confirms order for Civaux 1; NDP surprise victor in Ontario; Czechoslovakia joins Foratom; Poland abandons Zarnowiec; Cogema's UP3 facility operationa; Wismut to clsoe mining operations; court rules in favor of US government in Yucca Mountain waste site; Italian government ratifies ENEL reactor shutdowns; and Seabrook enters commercial operation

  11. Predicted irradiation behavior of U3O8-Al dispersion fuels for production reactor applications

    International Nuclear Information System (INIS)

    Cronenberg, A.W.; Rest, J.

    1990-01-01

    Candidate fuels for the new heavy-water production reactor include uranium/aluminum alloy and U 3 O 8 -Al dispersion fuels. The U 3 O 8 -Al dispersion fuel would make possible higher uranium loadings and would facilitate uranium recycle. Research efforts on U 3 O 8 -Al fuel include in-pile irradiation studies and development of analytical tools to characterize the behavior of dispersion fuels at high-burnup. In this paper the irradiation performance of U 3 O 8 -Al is assessed using the mechanistic Dispersion Analysis Research Tool (DART) code. Predictions of fuel swelling and alteration of thermal conductivity are presented and compared with experimental data. Calculational results indicate good agreement with available data where the effects of as-fabricated porosity and U 3 O 8 -Al oxygen exchange reactions are shown to exert a controlling influence on irradiation behavior. The DART code is judged to be a useful tool for assessing U 3 O 8 -Al performance over a wide range of irradiation conditions

  12. European Fast Reactor IWGFR/FRCC-report. A review of the collaborative programme on the European Fast Reactor (EFR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

    The design work for the 1500 MWe European Fast Reactor EFR was started in 1988. Two years during phase 1 were devoted to the concept design; the subsequent concept validation phase 2 will last until March 1993. In autumn 1991 the 'concept design '91, CD91, was put forward; its major design features and the R and D support are described briefly together with the organisational structures. The European Fast Reactor Utilities Group 'EFRUG' presently comprises EdF (France), ENEL (Italy), Nuclear Electric (UK) and Bayernwerk, PreulsenElektra and RWE (Germany). For design and construction of EFR the group 'EFR Associates' is responsible, combining the companies Siemens (formerly Interatom, Germany), NNC Ltd. (UK) and Framatome/Div. Novatome (France). The necessary R and D support is given by CEA (France), UKAEA (UK) and KfK/Siemens (Germany). The R and D work is executed in the various national research centres ranging from Dounreay via Bensberg and Karlsruhe to Cadarache. The design work is done at Bensberg, Lyon and Risley. The present programme of design work extends to early 1993 and is aimed at producing a detailed consistent design for the nuclear part of the plant and a non site specific safety report. By that date the basic feasibility of the main design features will have been underwritten by the joint R and D programme and there will be an informal assessment of the general licensibility of the concept by the Ad Hoc Safety Club. In follow-up the utilities will then be in the position to decide whether to proceed with the next steps. The key issues of this phase will include the specification of the plant, the siting, the detailed engineering, licensing with possibly a public enquiry and the question of ownership and financing. In the international arena the collaboration with USSR is proceeding well on the basis of an USSR-Europe Agreement from January 8, 1991; it foresees review and specialists meetings in the field of fast breeder research. On the occasion of

  13. Fabrication of zero power reactor fuel elements containing 233U3O8 powder

    International Nuclear Information System (INIS)

    Nicol, R.G.; Parrott, J.R.; Krichinsky, A.M.; Box, W.D.; Martin, C.W.; Whitson, W.R.

    1982-05-01

    Oak Ridge National Laboratory, under contract with Argonne National Laboratory, completed the fabrication of 1743 fuel elements for use in their Zero Power Reactor. The contract also included recovery of 20 kg of 233 U from rejected elements. This report describes the steps associated with conversion of purified uranyl nitrate (as solution) to U 3 O 8 powder (suitable for fuel) and subsequent charging, sealing, decontamination, and testing of the fuel elements (packets) preparatory to shipment. The nuclear safety, radiation exposures, and quality assurance aspects of the program are discussed

  14. Nuclear cycle of thorium and molt salts reactors. PE 5.8

    International Nuclear Information System (INIS)

    Doubre, H.

    2004-01-01

    In the framework of the nuclear industry development, many scenario are studied from the standard reactors using enriched uranium to the IV generation reactors. The study of new systems for the future of the nuclear needs to develop new simulation tools. The research programs of the IPN of Orsay are presented. (A.L.B.)

  15. Transactions of the 8th International Conference on Structure Mechanics in Reactor Technology

    International Nuclear Information System (INIS)

    Browzin, B.S.

    1985-06-01

    These Transactions of the JK-panel session include preprints of papers or abstracts which are listed in Volume A, ''Introduction, General Contents, Authors Index,'' Proceedings of the 8th International Conference on Structural Mechanics in Reactor Technology. These papers represent the body of the JK-panel session, ''Status of Research in Structural and Mechanical Engineering for Nuclear Power Plants,'' sponsored by the US Nuclear Regulatory Commission. Additional papers are expected at this session, which will be available at the session. The purpose of publishing these Transactions is to inform the participants of the JK-panel session in advance on the papers to be presented and discussed at the session

  16. Commissioning of qualification of structures, systems and components for seismic and environmental loads of CIRENE nuclear power plant

    International Nuclear Information System (INIS)

    Bianchi, A.; Gatti, F.; Muzzi, F.; Zola, M.; De Pasquali, F.

    1993-01-01

    On behalf of the Italian National Electricity Board (ENEL) concerning the commissioning of qualification of structures, systems and components of CIRENE NPP, ISMES performed a technical surveillance on the documentation concerning the environmental and seismic qualification of the safety related systems and experimental activities (dynamic and static tests) on plant buildings. The aims of the work were: - the evaluation of the qualification carried out (by test, by analysis, by combination of analysis and test) on the equipment and system, compared with the requirements of the ENEL technical specifications and the most recent international regulations; - the experimental determination of modal quantities (frequencies, damping, mode shapes) of the structures and, in the case of reactor building, the complex impedance of the soil for supporting the analytical work. The present paper deals with the criteria, the system and the results concerning the technical surveillance and with the characteristics and the results of the experimental tests

  17. Calibration of RB reactor power

    International Nuclear Information System (INIS)

    Sotic, O.; Markovic, H.; Ninkovic, M.; Strugar, P.; Dimitrijevic, Z.; Takac, S.; Stefanovic, D.; Kocic, A.; Vranic, S.

    1976-09-01

    The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

  18. Project requirements for reconstruction of the RA reactor ventilation system, Task 2.8. Measurement of radioactive iodine and other isotopes contents in the gas system of the RA reactor, Annex of the task

    International Nuclear Information System (INIS)

    Vujisic, Lj. et al

    1981-01-01

    This report is a supplement to the task 2.8. When planning and constructing the ventilation system, it was found that it is necessary to perform additional experiments during RA reactor operation at 2 MW power level for a longer period. In addition to the helium system, the potential source of radioactive pollutants is the space below the upper water shielding of the reactor. All the experimental and fuel channels are ending in this space. During repair and fuel exchange radioactivity can be released in this space. For that reason this space is important when planing and designing the filtration system for incidental conditions or planned dehermetisation of the reactor. The third point where radioactive isotope identification was done, was the entrance into the chimney during steady state operation and planned dehermetisation of the reactor. The following samples were measured: gas system during reactor operation at 2 MW power; entrance into the chimney during last 48 hours of reactor operation at 2 MW power; sample on the platform under the upper water shield with the opened fuel channel after the reactor shutdown; and simultaneously with the latter, measurement at the entrance to the chimney. This annex contains the list of identified radioactive isotopes, volatile and gaseous as well as concentration of volatile 131 I on the adsorbents [sr

  19. Licensed operating reactors. Status summary report, data as of 8-31-82

    International Nuclear Information System (INIS)

    1982-09-01

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  20. Perspective of nuclear power policy change and trend of nuclear industry activities from energy policy of European countries

    International Nuclear Information System (INIS)

    Murakami, Tomoko; Matsuo, Yuji; Nagatomi, Yu

    2009-01-01

    European countries of nuclear power phase-out have changed to commit to the future of nuclear energy due to the intended low-carbon power, the energy security concerns and the need of replacement reactors as current reactors approach the end of operating lives, as Italian government has passed legislation to build new nuclear power plants. This article described the perspective of nuclear power policy changes in UK, Italy an Sweden and the business trend and the SWOT analysis of related electric utilities (EDF, Enel and Vattenfall) and nuclear industries (Areva NP, Sheffield Forgemasters, ENSA and Studsvik). Policy implications obtained from this analysis were commented for Japanese nuclear industry activities. (T. Tanaka)

  1. Analysis of gamma dose for 4,8 gU/cm3 density silicide core at the RSG-GAS reactor using MCNP code

    International Nuclear Information System (INIS)

    Ardani

    2011-01-01

    Radiation safety analysis should be done following of substitution of fuel density of 2.96 gU/cc to density of 4,8 gU/cc silicide fuels for the RSG-GAS reactor. MCNP-5 code has been used to perform gamma dose calculation of the RSG-GAS reactor. Gamma radiation source at reactor consists of capture gamma rays, prompt fission gamma rays, and gamma rays of decay of fission and activation products. The strength of the prompt fission gamma rays is obtained by gamma releases of fission process of U-235 and reactor power of 30 MWt., during 46,6 days operation. Radiation dose is calculated at the experimental hall by detection point at the surface of outer of biological shielding and the operation hall by detection point at the top of the pool. The calculation is conducted at reactor on the normal operation and on the worst postulated accident causing the water level at the pool decreases. Calculation result shows that the biggest source strength of gamma rays come from the decay process. The highest calculated dose at the experiment hall is 4,07x10 -3 μSv/h, far from the maximum external dose permitted 25 μSv/h. The highest calculated dose at the operation hall is 19.98 μSv/h. Even though the calculated dose is still acceptable but this is close to the maximum permitted dose for worker. It concluded that loading of 4,8 gU/cc silicide fuel for the RSG-GAS still safe. (author)

  2. Status and programme of development of the fast breeder reactor systems in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Daeunert, U.; Kessler, G.

    1977-01-01

    Owing to the dependence of the Federal Republic of Germany on imports of raw materials for energy purposes, special importance is given to nuclear energy for reasons of energy policy. Against the background of a nuclear power station capacity amounting to 24,000 MWe in 1976 (power stations being either on order, under construction or actually in operation), the Fast Breeder Reactor (FBR) has quite outstanding significance in this country where the economical utilization both of uranium and of the plutonium accumulating in light water reactors is concerned. Work in this sector started in the Federal Republic of Germany in 1960. Milestones on the path of development were the initial activities on the plutonium critical SNEAK at Karlsruhe, the successful operation of the KNK 20 MWe experimental reactor (sodium-cooling) with the KNK II (plutonium core) follow-up stage, the construction of the 300 MWe FBR prototype nuclear power station at Kalkar, as well as the initial planning work carried out in connection with a 1,300 MWe FBR demonstration nuclear power station. The provision of a viable infrastructure for safety assessment must go hand in hand with the development of new reactor systems. For this reason, prototype reactors in the Federal Republic of Germany are already subjected to the same strict licensing procedures under nuclear legislation as are commercial nuclear power stations. After several years' practical experience, it is becoming evident that this approach was the right one to take. Already early on, it was recognized in the Federal Republic of Germany that the execution of long-term technological major projects were increasingly exceeding the capacities of unaided development by single nations. The SNR 300 FBR prototype project combines the efforts of the governments, power supply companies, reactor manufacturers and nuclear research centres of the Federal Republic of Germany and of the Kingdoms of both Belgium and the Netherlands. The Agreement signed

  3. Assessment of torsatrons as reactors

    International Nuclear Information System (INIS)

    Lyon, J.F.; Painter, S.L.

    1992-12-01

    Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R 0 = 6.6-8.8 m, on-axis magnetic field B 0 = 4.8-7.5 T, B max (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions

  4. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    Veeraraghaven, N.

    1990-01-01

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  5. Impact of proposed research reactor standards on reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

    1974-07-01

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  6. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

    Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

    1974-01-01

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  7. Nuclear Reactor RA Safety Report, Vol. 8, Auxiliary system; Izvestaj o sigurnosti nuklearnog reaktora RA, Knjiga 8, Pomocni sistemi

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-11-01

    This volume describes RA reactor auxiliary systems, as follows: special ventilation system, special drainage system, hot cells, systems for internal transport. Ventilation system is considered as part of the reactor safety and protection system. Its role is eliminate possible radioactive particles dispersion in the environment. Special drainage system includes pipes and reservoirs with the safety role, meaning absorption or storage of possible radioactive waste water from the reactor building. Hot cells existing in the RA reactor building are designed for production of sealed radioactive sources, including packaging and transport. [Serbo-Croat] Ova knjiga obuhvata opis pomocnih sistem reaktora RA: sistem specijalne ventilacije, sistem specijalne kanalizacije, vruce komore, sistemi za unutrasnji transport. Ventilacioni sistem je znacajan deo sistema zastite i sigurnosti reaktora. Njegova je uloga da onemoguci disperziju radioaktivnih cestica u okolinu. Sistem specijalne kanalizacije sastoji se od agregata, cevovoda i rezervoara sa sigurnosnom armaturom i ima zadatak da prihvata i u sebe deponuje radioaktivne otpadne vode iz objekarta reaktora RA. U zgradi reaktora RA postoje vruce komore koje su namenjene za proizvodnju zatvorenih izvora zracenje, ukljucujuci i mehanicku obradu, prepakivanje i transport.

  8. A comparison of the free vacancy production in α brass by fission reactor neutrons and 14.8-MeV neutrons

    International Nuclear Information System (INIS)

    Damask, A.C.; Van Konynenburg, R.; Borg, R.J.; Dienes, G.J.

    1976-01-01

    Enhancement of substitutional diffusion is observed in α brass (30 wt% Zn) by following the decrease in electrical resistivity with neutron irradiation of a thermally equilibrated alloy; the decrease arises from the increase in short-range order. It was determined by previous research that this diffusion enhancement is largely caused by the annealling of radiation-produced vacancies in excess of the thermal equilibrium concentration. Therefore, the results reported here are based upon a well-established technique. The rate of resistivity change per neutron of different energies will give the relative number of free vacancies produced per neutron. This experiment compares the effect of 14.8 MeV neutrons with neutrons from a fission reactor. The results indicate that 14.8 MeV neutrons produce 10 +- 2 times as many free vacancies as reactor neutrons when the latter are expressed in terms of those neutrons with energies greater than 0.1 MeV. (author)

  9. Compact torsatron reactors

    International Nuclear Information System (INIS)

    Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

    1988-05-01

    Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R 0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R 0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

  10. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

    International Nuclear Information System (INIS)

    1986-11-01

    RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

  11. Conceptual design of reactor assembly of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Selvaraj, A.; Balasubramaniyan, V.; Raghupathy, S.; Elango, D.; Sodhi, B.S.; Chetal, S.C.; Bhoje, S.B.

    1996-01-01

    The conceptual design of Reactor Assembly of 500 MWe Prototype Fast Breeder Reactor (as selected in 1985) was reviewed with the aim of 'simplification of design', 'Compactness of the reactor assembly' and 'ease in construction'. The reduction in size has been possible by incorporating concentric core arrangement, adoption of elastomer seals for Rotatable plugs, fuel handling with one transfer arm type mechanism, incorporation of mechanical sealing arrangement for IHX at the penetration in Inner vessel redan and reduction in number of components. The erection of the components has been made easier by adopting 'hanging' support for roof slab with associated changes in the safety vessel design. This paper presents the conceptual design of the reactor assembly components. (author). 8 figs, 2 tabs

  12. Magnetite synthesis from ferrous iron solution at pH 6.8 in a continuous stirred tank reactor.

    Science.gov (United States)

    Mos, Yvonne M; Zorzano, Karin Bertens; Buisman, Cees J N; Weijma, Jan

    2018-04-01

    Partial oxidation of defined Fe 2+ solutions is a well-known method for magnetite synthesis in batch systems. The partial oxidation method could serve as basis for an iron removal process in drinking water production, yielding magnetite (Fe 3 O 4 ) as a compact and valuable product. As a first step toward such a process, a series of experiments was carried out, in which magnetite was synthesized from an Fe 2+ solution in a 2 L continuous stirred tank reactor (CSTR) at atmospheric pressure and 32 °C. In four experiments, elevating the pH from an initial value of 5.5 or 6.0 to a final value of 6.8, 7.0 or 7.5 caused green rust to form, eventually leading to magnetite. Formation of NH 4 + in the reactor indicated that NO 3 - and subsequently NO 2 - served as the oxidant. However, mass flow analysis revealed an influx of O 2 to the reactor. In a subsequent experiment, magnetite formation was achieved in the absence of added nitrate. In another experiment, seeding with magnetite particles led to additional magnetite precipitation without the need for a pH elevation step. Our results show, for the first time, that continuous magnetite formation from an Fe 2+ solution is possible under mild conditions, without the need for extensive addition of chemicals.

  13. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    Jaouen, C.; Beroux, P.

    2012-01-01

    From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

  14. Ninth meeting of the International Working Group on Gas Cooled Reactors, Oak Ridge, USA, 8-9 November 1990

    International Nuclear Information System (INIS)

    1991-05-01

    This report contains the minutes of the meeting, the papers presented as overview of the national programmes in the field of gas-cooled reactors and the main results from discussions on the different items of the agenda. The meeting was attended by 20 members and/or alternates from 9 countries and 2 international organizations. 8 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  15. Study of the U3O8-Al thermite reaction and strength of reactor fuel tubes

    International Nuclear Information System (INIS)

    Peacock, H.B.

    1983-08-01

    Heating tests using 53 wt % U 3 O 8 -Al pellets show that an exothermic reaction occurs between 875 and 1000 0 C and takes 10 to 20 seconds to reach maximum temperature. The maximum temperature is a function of particle size of the U 3 O 8 with large particles exhibiting lower peak temperatures. The calculated energy release was 123 cal/g of U 3 O 8 -aluminum fuel. Tests using aluminum clad outer fuel tube sections gave lower peak temperatures than for pellets. No violent reactions occurred. The results are reasonably consistent with recent reported data indicating that the exothermic U 3 O 8 -Al reaction is not an important energy source. The compressive and tensile strengths of U 3 O 8 tubes above 660 0 C are low. In compression, sections with 2 psi average axial stress failed at 917 0 C, while sections with 7 psi failed at 669 0 C. Tubes with U-Al alloy cores failed at about 670 0 C with no applied load. The stresses in fuel tubes during a reactor transient may range up to several hundred psi and are less than 7 psi only in the upper part of the fuel tube

  16. Reactor Neutron Sources

    International Nuclear Information System (INIS)

    Aksenov, V.L.

    1994-01-01

    The present status and the prospects for development of reactor neutron sources for neutron scattering research in the world are considered. The fields of application of neutron scattering relative to synchrotron radiation, the creation stages of reactors (steady state and pulsed) and their position in comparison with spallation neutron sources at present and in the foreseen future are discussed. (author). 15 refs.; 8 figs.; 3 tabs

  17. Fusion reactor materials

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The following topics are briefly discussed: (1) surface blistering studies on fusion reactor materials, (2) TFTR design support activities, (3) analysis of samples bombarded in-situ in PLT, (4) chemical sputtering effects, (5) modeling of surface behavior, (6) ion migration in glow discharge tube cathodes, (7) alloy development for irradiation performance, (8) dosimetry and damage analysis, and (9) development of tritium migration in fusion devices and reactors

  18. Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J

    1972-07-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

  19. Advances in Reactor Physics, Mathematics and Computation. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.

  20. Annual report on JEN-1 reactor

    International Nuclear Information System (INIS)

    Montes, J.

    1972-01-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

  1. ITER [International Thermonuclear Experimental Reactor] reactor building design study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Blevins, J.D.; Delisle, M.W.

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is at the midpoint of a two-year conceptual design. The ITER reactor building is a reinforced concrete structure that houses the tokamak and associated equipment and systems and forms a barrier between the tokamak and the external environment. It provides radiation shielding and controls the release of radioactive materials to the environment during both routine operations and accidents. The building protects the tokamak from external events, such as earthquakes or aircraft strikes. The reactor building requirements have been developed from the component designs and the preliminary safety analysis. The equipment requirements, tritium confinement, and biological shielding have been studied. The building design in progress requires continuous iteraction with the component and system designs and with the safety analysis. 8 figs

  2. Status of French reactors

    International Nuclear Information System (INIS)

    Ballagny, A.

    1997-01-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm 3 . The OSIRIS reactor has already been converted to LEU. It will use U 3 Si 2 as soon as its present stock of UO 2 fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU

  3. Reactor-specific spent fuel discharge projections, 1987-2020

    International Nuclear Information System (INIS)

    Walling, R.C.; Heeb, C.M.; Purcell, W.L.

    1988-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs

  4. Extension of cycle 8 of Angra-1 reactor, optimization of electric power generation reduction

    International Nuclear Information System (INIS)

    Miranda, Anselmo Ferreira; Moreira, Francisco Jose; Valladares, Gastao Lommez

    2000-01-01

    The main objective of extending fuel cycle length of Angra-1 reactor, is in fact of that each normal refueling are changed about 40 fuel elements of the reactor core. Considering that these elements do not return for the reactor core, this procedure has became possible a more gain of energy of these elements. The extension consists in, after power generation corresponding to a cycle burnup of 13700 MWD/TMU or 363.3 days, to use the reactivity gain by reduction of power and temperature of primary system for power generation in a low energy patamar

  5. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Martinc, R.; Cupac, S.; Stanic, A.

    1990-01-01

    RA reactor was not operated during the past five years due to the renewal and reconstruction of the reactor systems, which in underway. In the period from 1986-1990, reactor was operated only 144 MWh in 1986, for the need of testing the reactor systems and possibility of irradiating 125 I. Reactor will not be operated in 1991 because of the exchange of complete instrumentation which is planned to be finished by the end of 1991. It is expected to start operation in May 1992. That is why this annex includes the plan of reactor operation for period of nine months starting from from the moment of start-up. It is planned to operate the reactor at 0.02 MW power first three months, to increase the power gradually and reach 3.5 MW after 8 months of operation. It is foreseen to operate the reactor at 4.7 MW from the tenth month on [sr

  6. Europe: a new start for the nuclear energy. BIP-Enerpresse Debate - February 10, 2009

    International Nuclear Information System (INIS)

    2009-01-01

    This paper contains a synthesis of a debate organised in the context of a gas supply crisis which leads some countries to launch nuclear reactor projects (such as in the United Kingdom) or to carry on their program (France, China). The participants, all representing important French actors (Areva, EDF, GDF Suez, Enel France), are commenting the statement of the French government announcing the construction of a second EPR in France, the evolution of the European institutional context, the legal environment in the United Kingdom and in Italy, the competitiveness and investment issues. They also answer questions about Areva's activity in Europe, the fourth generation reactors, the needs for skilled engineers, technicians and workers, the new start of nuclear energy in Italy

  7. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  8. Molten salt reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Simon, N.; Renault, C.

    2014-01-01

    Molten salt reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. The principle of this reactor is very innovative: the nuclear fuel is dissolved in the coolant which allows the online reprocessing of the fuel and the online recovery of the fission products. A small prototype: the Molten Salt Reactor Experiment (MSRE - 8 MWt) was operating a few years in the sixties in the USA. The passage towards a fast reactor by the suppression of the graphite moderator leads to the concept of Molten Salt Fast Reactor (MSFR) which is presently studied through different European projects such as MOST, ALISIA and EVOL. Worldwide the main topics of research are: the adequate materials resisting to the high level of corrosiveness of the molten salts, fuel salt reprocessing, the 3-side coupling between neutron transport, thermohydraulics and thermo-chemistry, the management of the changing chemical composition of the salt, the enrichment of lithium with Li 7 in the case of the use of lithium fluoride salt and the use of MSFR using U 233 fuel (thorium cycle). The last part of the article presents a preliminary safety analysis of the MSFR. (A.C.)

  9. RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1987-01-01

    RA research reacto was not operated due to the prohibition issued in 1984 by the Government of Serbia. Three major tasks were finished in order to fulfill the licensing regulations about safety of nuclear facilities which is the condition for obtaining permanent operation licence. These projects involved construction of the emergency cooling system, reconstruction of the existing special ventilation system, and renewal of the system for electric power supply of the reactor systems. Renewal of the RA reactor instrumentation system was initiated. Design project was done by the Russian Atomenergoeksport, and is foreseen to be completed by the end of 1988. The RA reactor safety report was finished in 1987. This annual report includes 8 annexes concerning reactor operation, activities of services and financial issues, and three special annexes: report on testing the emergency cooling system, report on renewal of the RA reactor and design specifications for reactor renewal and reconstruction [sr

  10. Economies of scale versus economies of density in the electricity distribution. A quantitative analysis; Economie di scala versus economie di densita' nella distribuzione elettrica. Un'analisi quantitativa

    Energy Technology Data Exchange (ETDEWEB)

    Gulli' , F. [IEFE, Universita' Bocconi, Milan (Italy)

    2000-06-01

    The aim of this article is to analyse the costs of the electricity distribution in Italy. The case examined (Enel's distribution activity) is particularly interesting. In fact the specific organization of Enel allows to separate the effects of firm scale from the influence of the customers territorial concentration (territorial density). On this subject, the results of the econometric analysis shows that density economics are more important than scale economics. [Italian] L'obiettivo di questo articolo e' analizzare i costi nell'attivita' di distribuzione elettrica in Italia. Il caso esaminato, la distribuzione dell'Enel, e' a questo proposito particolarmente interessante, grazie alla specifica organizzazione dell'Enel che prevede la presenza di centri di gestione gerarchicamente separati. E' stato cosi' possibile distinguere nettamente fra effetti dovuti alla scala produttiva ed effetti connessi alla concentrazione territoriale dell'utenza (effetti di densita'). Il modello econometrico costruito sui dati dell'Enel ha, al riguardo, permesso di verificare che le economie di densita' sono piu' importanti di quelle di scala.

  11. RB Research nuclear reactor, Annual report for 2004

    International Nuclear Information System (INIS)

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

    2005-02-01

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

  12. Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

    Energy Technology Data Exchange (ETDEWEB)

    Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Tsai Hanchung [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439-4803 (United States)

    2012-08-15

    The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 Degree-Sign C, cooling to 522 Degree-Sign C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at {approx}0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher {Delta}T between fuel center and cladding than at core center, together providing more rare earths at the cladding and

  13. Reactor-specific spent fuel discharge projections: 1986 to 2020

    International Nuclear Information System (INIS)

    Heeb, C.M.; Walling, R.C.; Purcell, W.L.

    1987-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No new orders with extended burnup, (2) No new orders with constant burnup, (3) Upper reference (which assumes extended burnup), (4) Upper reference with constant burnup, and (5) Lower reference (which assumes extended burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel. 6 refs., 8 figs., 8 tabs

  14. Calibration of RB reactor power; Kalibrisanje snage reaktora RB

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Markovic, H; Ninkovic, M; Strugar, P; Dimitrijevic, Z; Takac, S; Stefanovic, D; Kocic, A; Vranic, S [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1976-09-15

    The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8{radical}2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation.

  15. Creep behavior of 8Cr2WVTa martensitic steel designed for fusion DEMO reactor. An assessment on helium embrittlement resistance

    International Nuclear Information System (INIS)

    Yamamoto, Norikazu; Murase, Yoshiharu; Nagakawa, Johsei; Shiba, Kiyoyuki

    2001-01-01

    Mechanical response against transmutational helium production, alternatively susceptibility to helium embrittlement, in a nuclear fusion reactor was examined on 8Cr2WVTa martensitic steel, a prominent structural candidate for advanced fusion systems. In order to simulate DEMO (demonstrative) reactor environments, helium was implanted into the material at 823 K with concentrations up to 1000 appmHe utilizing an α-beam from a cyclotron. Creep rupture properties were subsequently determined at the same temperature and were compared with those of the material without helium. It has been proved that helium caused no meaningful deterioration in terms of both the creep lifetime and rupture elongation. Furthermore, failure occurred completely in a transgranular and ductile manner even after high concentration helium introduction and there was no symptom of grain boundary decohesion which very often arises in helium bearing materials. These facts would mirror preferable resistance of this steel toward helium embrittlement. (author)

  16. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1986

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1986-01-01

    In order to enable future reliable operation of the RA reactor, according to new licensing regulations, three major tasks started in 1984 were fulfilled: building of the new emergency system, reconstruction of the existing ventilation system, and reconstruction of the power supply system. Simultaneously in 1985/1986 renewal of the instrumentation and reconstruction of the system for handling and storage of the spent fuel in the reactor building have started. Design projects for these tasks are almost finished and the reconstruction of both systems is expected to be finished until 1988 and mid 1989 respectively. RA reactor Safety report was finished according to the recommendations of the IAEA. Investments in 1986 were used for 8000 kg of heavy water, maintenance of reactor systems and supply of new components, reconstruction of reactor systems. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

  17. Measuring set: Reactor Power Meter (type of SG-8), Reactor Energy Meter (type of SG-11) and Digital Dose Meter (type of SG-9) for reactor rigs operation. Zestaw pomiarowy: miernik mocy reaktora (typ SG-8), miernik energii reaktora (typ SG-11) oraz cyfrowy miernik dawki (typ SG-9) dla potrzeb eksploatacji sond reaktorowych

    Energy Technology Data Exchange (ETDEWEB)

    Glowacki, S W

    1982-01-01

    A measuring set consisting of the Reactor Power Meter, Reactor Energy Meter and Digital Dose Meter is described. The gamma radiation of water in the reactor primary cooling circuit reaches the ionisation chamber and involves the output current, driving the Reactor Power Meter and Reactor Energy Meter. The Digital Dose Meter is controlled by the output current of the self-powered detector mounted inside the reactor rig.

  18. Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

    Energy Technology Data Exchange (ETDEWEB)

    Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science in Radiation and Dynamics Extremes

    2016-09-26

    In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

  19. Evaluation of VVER-1200/V-491 reactor pressure vessel integrity during large break LOCA along with SBO using MELCOR 1.8.6

    International Nuclear Information System (INIS)

    Bui Thi Hoa; Tran Chi Thanh

    2015-01-01

    After Fukushima accident and stress test recommended by IAEA for existing reactors, higher safety requirements are enforced upon nuclear power plants during design extension and severe accident conditions. Based on those arguments, Vietnam Government requests a lot of effective safety solutions, in designs proposed for the nuclear power plants in Ninh Thuan province of Vietnam, which can prevent the accident progression toward severe accidents and mitigate severe accident consequences. One of safety requirements is related to delay time of core melt during design extension condition. Especially, if the worst case of accidents occurs, the reactor vessel integrity must be maintained at least 24 hours from the beginning of the accident. With the aim at investigation of Reactor Pressure Vessel (RPV) integrity, in this study, MELCOR 1.8.6 code is used to evaluate the integrity of RPV lower head for VVER-1200/V-491 reactor during a Large Break Loss of Coolant Accident (LBLOCA) in combination with Station Blackout (SBO) event. The study figures out several parameters related to melt down progress such as: rupture position and rupture timing, the amount of hydrogen generated. Availability of the second stage hydro-accumulators (HA2) in the VVER-1200/V-491 is assumed as an additional improvement to delay the timing of core melt as well as to maintain the vessel integrity for long-term. (author)

  20. Fusion reactors as a future energy source

    International Nuclear Information System (INIS)

    Seifritz, W.

    A detailed update of fusion research concepts is given. Discussions are given for the following areas: (1) the magnetic confinement principle, (2) UWMAK I: conceptual design for a fusion reactor, (3) the inertial confinement principle, (4) the laser fusion power plant, (5) electron-induced fusion, (6) the long-term development potential of fusion reactors, (7) the symbiosis between fusion and fission reactors, (8) fuel supply for fusion reactors, (9) safety and environmental impact, and (10) accidents, and (11) waste removal and storage

  1. Oscillating liquid flow ICF Reactor

    International Nuclear Information System (INIS)

    Petzoldt, R.W.

    1990-01-01

    Oscillating liquid flow in a falling molten salt inertial confinement fusion reactor is predicted to rapidly clear driver beam paths of residual liquid droplets. Oscillating flow will also provide adequate neutron and x-ray protection for the reactor structure with a short (2-m) fall distance permitting an 8 Hz repetition rate. A reactor chamber configuration is presented with specific features to clear the entire heavy-ion beam path of splashed molten salt. The structural components, including the structure between beam ports, are shielded. 3 refs., 12 figs

  2. Proceedings of the international topical seminar on management of ageing of research reactors, Geesthacht/Hamburg, May 8-12, 1995

    International Nuclear Information System (INIS)

    Alcala-Ruiz, F.; Krull, W.

    1995-01-01

    The GKSS Research Centre Geesthacht GmbH and the International Atomic Energy Agency (IAEA) held a joint seminar on 'Management of Ageing of Research Reactors' from May 8-12, 1995, at Geesthacht/Hamburg. More than 100 participants from 36 countries and two international organizations were present. During the seminar 52 papers have been presented discussing the large variety of ageing effects from physical ageing to design ageing and staff ageing and the measures taken to identify and to overcome the ageing effects. (orig.) [de

  3. Managing severe reactor accidents. A review and evaluation of our knowledge on reactor accidents and accident management

    International Nuclear Information System (INIS)

    Gustavsson, Veine

    2002-11-01

    The report gives a review of the results from the last years research on severe reactor accidents, and an opinion on the possibilities to refine the present strategies for accident management in Swedish and Finnish BWRs. The following aspect of reactor accidents are the major themes of the study: 1. Early pressure relief from hydrogen production; 2. Recriticality in re-flooded, degraded core; 3. Melt-through; 4. Steam explosion after melt-through; 5. Coolability of the melt after after melt-through; 6. Hydrogen fire in the reactor containment; 7. Leaking containment; 8. Hydrogen fire in the reactor building; 9. Long-time developments after a severe accident; 10. Accidents during shutdown for overhaul; 11. Information need for remedial actions. Possibilities for improving the strategies in each of these areas are discussed. The review shows that our knowledge is sufficient in the areas 1, 2, 4, 6, 8. For the other areas, more research is needed

  4. Effect of temperature on selenium removal from wastewater by UASB reactors.

    Science.gov (United States)

    Dessì, Paolo; Jain, Rohan; Singh, Satyendra; Seder-Colomina, Marina; van Hullebusch, Eric D; Rene, Eldon R; Ahammad, Shaikh Ziauddin; Carucci, Alessandra; Lens, Piet N L

    2016-05-01

    The effect of temperature on selenium (Se) removal by upflow anaerobic sludge blanket (UASB) reactors treating selenate and nitrate containing wastewater was investigated by comparing the performance of a thermophilic (55 °C) versus a mesophilic (30 °C) UASB reactor. When only selenate (50 μM) was fed to the UASB reactors (pH 7.3; hydraulic retention time 8 h) with excess electron donor (lactate at 1.38 mM corresponding to an organic loading rate of 0.5 g COD L(-1) d(-1)), the thermophilic UASB reactor achieved a higher total Se removal efficiency (94.4 ± 2.4%) than the mesophilic UASB reactor (82.0 ± 3.8%). When 5000 μM nitrate was further added to the influent, total Se removal was again better under thermophilic (70.1 ± 6.6%) when compared to mesophilic (43.6 ± 8.8%) conditions. The higher total effluent Se concentration in the mesophilic UASB reactor was due to the higher concentrations of biogenic elemental Se nanoparticles (BioSeNPs). The shape of the BioSeNPs observed in both UASB reactors was different: nanospheres and nanorods, respectively, in the mesophilic and thermophilic UASB reactors. Microbial community analysis showed the presence of selenate respirers as well as denitrifying microorganisms. Copyright © 2016 Elsevier Ltd. All rights reserved.

  5. Vibrations measurement in fast and PWR reactor study

    International Nuclear Information System (INIS)

    Tigeot, Y.; Epstein, A.; Hareux, F.

    1975-01-01

    In the past severe damages have occured in several nuclear reactors, by structural vibrations induced by the primary cooling flow. To avoid this kind of troubles, the SEMT makes studies for two different types of reactors. For the light pressurized water reactors, some tests have been made on the SAFRAN test loop which is a three loop 1/8 scale internal model of a 900 MWe reactor. This study is actually undertaken jointly with Framatome. Elsewhere, measurements have been made on the Phenix fast breeder sodium reactor, and studies are planned for the Super Phenix reactor [fr

  6. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Carlos Ruben Calabrese

    1999-01-01

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  7. Dose rate in the reactor room and environment during maintenance in fusion reactors

    International Nuclear Information System (INIS)

    Maki, Koichi; Satoh, Satoshi; Takatsu, Hideyuki; Seki, Yasushi

    1995-01-01

    According to the International Thermonuclear Experimental Reactor (ITER) conceptual design activity, after reactor shutdown, damaged segments are pulled up from the reactor and hung from the reactor room ceiling by a remote handling device. The dose rate in the reactor room and the environment is estimated for this situation, and the following results are obtained. First, the dose rate in the room is > 10 8 μSv/h. Since this dose rate is 10 7 times greater than the biological radiation shielding design limit of 25 μSv/h, workers cannot enter the room. Second, lenses and optical fiber composed of glass that is radiation resistant up to 10 6 Gy would be damaged after <100 h near the segment, and devices using semiconductors could not work after several hours or so in the aforementioned dose-rate conditions. Third, during suspension of one blanket segment from the ceiling, the dose rate in the site boundary can be reduced by one order by a 23-cm-thicker reactor building roof. To reduce dose rate in public exposure to a value that is less than one-tenth of the public exposure radiation shielding design limit of 100 μSv/yr, the distance of the site boundary from the reactor must be greater than 200 m for a reactor building with a 160-cm-thick concrete roof. 9 refs., 6 figs., 2 tabs

  8. Power Trip Set-points of Reactor Protection System for New Research Reactor

    International Nuclear Information System (INIS)

    Lee, Byeonghee; Yang, Soohyung

    2013-01-01

    This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

  9. Specialists' meeting on heat and mass transfer in the reactor cover gas, Harwell, England, 8-10 October 1985

    International Nuclear Information System (INIS)

    1986-07-01

    The specialists' meeting on ''Heat and Mass Transfer in the Reactor Cover Gas'' was held at Harwell, the United Kingdom, on 8-10 October 1985. It was attended by 24 participants from all IWGFR member-countries: France, the Federal Republic of Germany, India, Italy, Japan, the Union of Soviet Socialist Republics, the United Kingdom and the United States. The meeting was presided over by Dr K. Eickhoff of the United Kingdom. The following topical areas were reviewed and discussed during the meeting: 1. National review presentations on the status of activities on heat and mass transfer in the reactor cover gas - 2 papers; 2. Aerosol dynamics - 4 papers; 3. Aerosol trapping - 2 papers; 4. Heat and mass transfer through cover gas in annuli - 3 papers; 5. Radiative properties - 4 papers; 6. Modelling of cover gas - 4 papers. A separate abstract was prepared for each of these papers. On the basis of papers presented and discussed by participants, session summaries and conclusions were drafted on the above topical areas. These summaries, as well as general conclusions and recommendations of the meeting were reviewed and agreed upon by consensus at the end of the meeting

  10. CANDU reactors with reactor grade plutonium/thorium carbide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

    2011-08-15

    Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

  11. Compact stellarators as reactors

    International Nuclear Information System (INIS)

    Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

    2001-01-01

    Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

  12. Practical course on reactor instrumentation

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2004-06-01

    This course is based on the description of the instrumentation of the TRIGA-reactor Vienna, which is used for training research and isotope production. It comprises the following chapters: 1. instrumentation, 2. calibration of the nuclear channels, 3. rod drop time of the control rods, 4. neutron flux density measurements using compensated ionization, 5. neutron flux density measurement with fission chambers (FC), 6. neutron flux density measurement with self-powered neutron detectors (SPND), 7. pressurized water reactor simulator, 8. verification of the radiation level during reactor operation. There is one appendix about neutron-sensitive thermocouples. (nevyjel)

  13. Analysis for the coolability of the reactor cavity in a Korean 1000 MWe PWR using MELCOR 1.8.3 computer code

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Kim, Ju Yeul; Chung, Chang Hyun; Park, Soo Yong

    1996-01-01

    The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction (MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass. The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment

  14. Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

    International Nuclear Information System (INIS)

    Ipek, O.

    1997-01-01

    The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO 2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S 16 -P 3 approach and DLC36 data library are used

  15. Molten-salt converter reactors

    International Nuclear Information System (INIS)

    Perry, A.M.

    1975-01-01

    Molten-salt reactors appear to have substantial promise as advanced converters. Conversion ratios of 0.85 to 0.9 should be attainable with favourable fuel cycle costs, with 235 U valued at $12/g. An increase in 235 U value by a factor of two or three ($10 to $30/lb. U 3 O 8 , $75/SWU) would be expected to increase the optimum conversion ratio, but this has not been analyzed in detail. The processing necessary to recover uranium from the fuel salt has been partially demonstrated in the MSRE. The equipment for doing this would be located at the reactor, and there would be no reliance on an established recycle industry. Processing costs are expected to be quite low, and fuel cycle optimization depends primarily on inventory and burnup or replacement costs for the fuel and for the carrier salt. Significant development problems remain to be resolved for molten-salt reactors, notably the control of tritium and the elimination of intergranular cracking of Hastelloy-N in contact with tellurium. However, these problems appear to be amenable to solution. It is appropriate to consider separating the development schedule for molten-salt reactors from that for the processing technology required for breeding. The Molten-Salt Converter Reactor should be a useful reactor in its own right and would be an advance towards the achievement of true breeding in thermal reactors. (author)

  16. Fast breeder reactor-block antiseismic design and verification

    International Nuclear Information System (INIS)

    Martelli, A.; Forni, M.

    1988-02-01

    The Specialists' Meeting on ''Fast Breeder Reactor-Block Antiseismic Design and Verification'' was organized by the ENEA Fast Reactor Department in co-operation with the International Working Group (IWGFR) of the International Atomic Energy Agency (IAEA), according to the recommendations of the 19th IAEA/IWGFR Meeting. It was held in Bologna, at the Headquarters of the ENEA Fast Reactor Department, on October 12-15, 1987, in the framework of the Celebrations for the Ninth Centenary of the Bologna University. The proceedings of the meeting consists of three parts. Part 1 contains the introduction and general comments, the agenda of the meeting, session summaries, conclusions and recommendations and the list of participants. Part 2 contains 8 status reports of Member States participating in the Working Group. Contributed papers were published in Part 3 and were further subdivided into 5 sessions as follows: whole reactor-block analysis (4 papers); whole reactor-block analysis (sloshing and buckling, seismic isolation effects) (8 papers); detailed core analysis (6 papers); shutdown systems and core structural and functional verifications (6 papers); component and piping analysis (7 papers). A separate abstract was prepared for each of the 8 status reports and 31 contributed papers. Refs, figs and tabs

  17. Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-02-01

    An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U 3 O 8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U 3 O 8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author) [pt

  18. RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1984

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1984-12-01

    During the 1984 the reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981. Operation of the primary cooling system was changed in order to avoid appearance of the previously noticed aluminium oxyhydrate on the surface of the fuel element claddings. The new cooling regime enabled more efficient heavy water purification. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks are planned: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. Financing of the planned activities will be partly covered by the IAEA. this Part I of the report includes 8 Annexes describing in detail the reactor operation, and 6 special papers dealing with the problems of reactor operation and utilization

  19. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1988-01-01

    According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

  20. Reactor Neutrino Oscillations: KamLAND and KASKA

    International Nuclear Information System (INIS)

    Suekane, F.

    2006-01-01

    Nuclear reactors generate a huge number of low energy ν-bar e 's. The reactor neutrinos have been used to study properties of neutrinos since its discovery a half century ago. Recently, KamLAND group finally discovered reactor neutrino oscillation with average baseline 180 km. According to the 3 flavor scheme of standard theory and measured oscillation parameters so far, the reactor neutrino is expected to perform another type of small oscillation at a baseline 1.8 km. KASKA experiment is a project to detect this small oscillation and to measure the last neutrino mixing angle θ 13 by using the world most powerful reactor complex, Kashiwazaki-Kariwa nuclear power station. In this proceedings, phenomena of neutrino oscillation and the two reactor oscillation experiments, KamLAND and KASKA, are introduced

  1. Bio-hydrogen production from molasses by anaerobic fermentation in continuous stirred tank reactor

    Science.gov (United States)

    Han, Wei; Li, Yong-feng; Chen, Hong; Deng, Jie-xuan; Yang, Chuan-ping

    2010-11-01

    A study of bio-hydrogen production was performed in a continuous flow anaerobic fermentation reactor (with an available volume of 5.4 L). The continuous stirred tank reactor (CSTR) for bio-hydrogen production was operated under the organic loading rates (OLR) of 8-32 kg COD/m3 reactor/d (COD: chemical oxygen demand) with molasses as the substrate. The maximum hydrogen production yield of 8.19 L/d was obtained in the reactor with the OLR increased from 8 kg COD/m3 reactor/d to 24 kg COD/m3 d. However, the hydrogen production and volatile fatty acids (VFAs) drastically decreased at an OLR of 32 kg COD/m3 reactor/d. Ethanoi, acetic, butyric and propionic were the main liquid fermentation products with the percentages of 31%, 24%, 20% and 18%, which formed the mixed-type fermentation.

  2. Licensed operating reactors. Status summary report as of February 29, 1984. Volume 8, No. 3

    International Nuclear Information System (INIS)

    1984-04-01

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Resource Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement (IE), from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  3. The zero power reactor SUR and its application

    International Nuclear Information System (INIS)

    Wesser, U.

    1986-01-01

    This low-power reactor, rated nominally at 100 milliwatts, has a cylindrical core of 26 cm in diameter and 24 cm high consisting of U 3 O 8 powder in a polyethylene matrix. The fuel is 20 percent enriched and the critical mass about 700 g. The excess reactivity is about 3 mk. The reactivity is controlled by two cadmium sheets in addition to a back-up system that drops the inner reflector. The reactor has no active cooling system. Personnel costs include a supervisor and an operator. The reactor is used for training in Reactor Theory (including use of a neutron chopper), reactor kinetics, nuclear technology, reactor operations and for doctoral thesis research. (author)

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

    International Nuclear Information System (INIS)

    1968-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1968-04-15

    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)

  6. LS1: electrical engineering upgrades and consolidation

    International Nuclear Information System (INIS)

    Duval, F.

    2012-01-01

    3 different types of activities are planned by the Engineering Department Electrical Engineering (EN-EL) Group for the first long shut down (LS1). First, the consolidation of EN-EL's ageing infrastructure elements. It is part of a 15- year programme aiming at increasing the reliability and availability of the power distribution network. Secondly, the maintenance of the accelerators infrastructure. In addition to the usual periodic operations and those delayed until LS1, the group will address more demanding activities like the replacement campaigns for irradiated cables and non-radiation resistant fibres as well as the removal of unused cables in particularly overcrowded areas. Thirdly, a vast amount of user copper and optical fibre cabling requests: EN-EL estimates that only 50% of LS1 requests are currently known. The main activities will be EN-EL's contributions to the R2E project, BE-BI upgrade projects, and the RF upgrade project in SPS BA3

  7. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    International Nuclear Information System (INIS)

    Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

    2014-01-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

  8. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

    2014-07-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

  9. Burnout experiments with 6 x 6, 8 x 8 and 7 x 7 rod bundle test sections using freon as model fluid

    International Nuclear Information System (INIS)

    Fulfs, H.; Katsaounis, A.; Minden, C.v.

    1976-01-01

    This paper reports on burnout experiments at staedy state condition using Freon12 as model fluid. The experiments were carried out with three test sections with 6 x 6, 8 x 8 and 7 x 7 rod bundles. The axial flux distribution of the rods is either constant or reactor like. The transformed measured points using STEVENS and BOURE scaling factors to equivalent water conditions respectively, were compared to the burnout correlation W3 using the reactor layout program DYNAMIT. The DYNAMIT code is a thermohydraulic lay-out reactor program without consideration of mixing flow between the subchannels. (orig.) [de

  10. Analysis of an accident type sbloca in reactor contention AP1000 with 8.0 Gothic code; Analisis de un accidente tipo Sbloca en la contencion del reactor AP1000 con el codigo Gothic 8.0

    Energy Technology Data Exchange (ETDEWEB)

    Goni, Z.; Jimenez Varas, G.; Fernandez, K.; Queral, C.; Montero, J.

    2016-08-01

    The analysis is based on the simulation of a Small Break Loss-of-Coolant-Accident in the AP1000 nuclear reactor using a Gothic 8.0 tri dimensional model created in the Science and Technology Group of Nuclear Fision Advanced Systems of the UPM. The SBLOCA has been simulated with TRACE 5.0 code. The main purpose of this work is the study of the thermo-hydraulic behaviour of the AP1000 containment during a SBLOCA. The transients simulated reveal close results to the realistic behaviour in case of an accident with similar characteristics. The pressure and temperature evolution enables the identification of the accident phases from the RCS point of view. Compared to the licensing calculations included in the AP1000 Safety Analysis, it has been proved that the average pressure and temperature evolution is similar, yet lower than the licensing calculations. However, the temperature and inventory distribution are significantly heterogeneous. (Author)

  11. SAP Nuclear Competence Centre

    International Nuclear Information System (INIS)

    Andrlova, Z.

    2009-01-01

    In this issue we continue and introduce the SAP Nuclear Competence Centre and its head Mr. Igor Dzama. SAP Nuclear Competence Centrum is one of the fi rst competence centres outside ENEL headquarters. It should operate in Slovakia and should have competencies within the whole Enel group. We are currently dealing with the issues of organisation and funding. We are trying to balance the accountability to the NPP directors and to the management of the competence centres at Enel headquarters; we are looking at the relations between the competence centres within the group and defining the services that we will provide for the NPPs. author)

  12. Introduction to Safety Analysis Approach for Research Reactors

    International Nuclear Information System (INIS)

    Park, Suki

    2016-01-01

    The research reactors have a wide variety in terms of thermal powers, coolants, moderators, reflectors, fuels, reactor tanks and pools, flow direction in the core, and the operating pressure and temperature of the cooling system. Around 110 research reactors have a thermal power greater than 1 MW. This paper introduces a general approach to safety analysis for research reactors and deals with the experience of safety analysis on a 10 MW research reactor with an open-pool and open-tank reactor and a downward flow in the reactor core during normal operation. The general approach to safety analysis for research reactors is described and the design features of a typical open-pool and open-tank type reactor are discussed. The representative events expected in research reactors are investigated. The reactor responses and the thermal hydraulic behavior to the events are presented and discussed. From the minimum CHFR and the maximum fuel temperature calculated, it is ensured that the fuel is not damaged in the step insertion of reactivity by 1.8 mk and the failure of all primary pumps for the reactor with a 10 MW thermal power and downward core flow

  13. Neutronic feasibility studies using U-Mo dispersion fuel (9 Wt % Mo, 5.0 gU/cm3) for LEU conversion of the MARIA (Poland), IR-8 (Russia), and WWR-SM (Uzbekistan) research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Deen, J.R.; Hanan, N.A.; Matos, E.

    2000-01-01

    U-Mo alloys dispersed in an Al matrix offer the potential for high-density uranium fuels needed for the LEU conversion of many research reactors. On-going fuel qualification tests by the US RERTR Program show good irradiation properties of U-Mo alloy dispersion fuel containing 7-10 weight percent molybdenum. For the neutronic studies in this paper the alloy was assumed to contain 9 wt % Mo (U-9Mo) with a uranium density in the fuel meat of 5.00 gU/cm 3 which corresponds to 32.5 volume % U-9Mo. Fuels containing U-9Mo have been used in Russian reactors since the 1950's. For the three research reactors analyzed here, LEU fuel element thicknesses are the same as those for the Russian-fabricated HEU reference fuel elements. Relative to the reference fuels containing 80-90% enriched uranium, LEU U-9Mo Al-dispersion fuel with 5.00 gU/cm 3 doubles the cycle length of the MARIA reactor and increases the IR-8 cycle length by about 11%. For the WWR-SM reactor, the cycle length, and thus the number of fuel assemblies used per year, is nearly unchanged. To match the cycle length of the 36% enriched fuel currently used in the WWR-SM reactor will require a uranium density in the LEU U-9Mo Al-dispersion fuel of about 5.4 gU/cm 3 . The 5.00 gU/cm 3 LEU fuel causes thermal neutron fluxes in water holes near the edge of the core to decrease by (6-8)% for all three reactors. (author)

  14. Guidelines for nuclear reactor equipments safety-analysis

    International Nuclear Information System (INIS)

    1978-01-01

    The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

  15. ''Sleeping reactor'' irradiations: Shutdown reactor determination of short-lived activation products

    International Nuclear Information System (INIS)

    Jerde, E.A.; Glasgow, D.C.

    1998-01-01

    At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux (φ) of ∼ 4 x 10 14 n/cm 2 · s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of ∼ 6 s, but the requirement of immediate counting leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about ± 0.5 s) make irradiations of 9 Be(γ,n) 8 Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to ∼ 1 x 10 10 n/cm 2 · s within 1 h. By the time the fuel elements are removed, the flux has dropped to ∼ 6 x 10 8 . Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant

  16. Preliminary shielding design evaluation for reactor assembly of SMART

    International Nuclear Information System (INIS)

    Kim, Kyo Youn; Kang, Chang M.; Kim, Ha Yong; Zee, Sung Quun; Chang, Moon Hee

    1999-03-01

    This report describes a preliminary evaluations of SMART shielding design near the reactor core by using the DORT two-dimensional discrete ordinates transport code. The results indicate that maximum neutron fluence at the bottom of reactor vessel is 1.64x10 17 n/cm 2 and that on the radial surface of reactor vessel is 6.71x10 16 n/cm 2 . These results meet the requirement, 1.0x10 20 n/cm 2 , in 10 CFR 50.61 and the integrity of SMART reactor vessel is confirmed during the lifetime of reactor. (Author). 20 refs., 11 tabs., 8 figs

  17. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  18. Power generation costs for alternate reactor fuel cycles

    International Nuclear Information System (INIS)

    Smolen, G.R.; Delene, J.G.

    1980-09-01

    The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

  19. Licensed operating reactors. Status summary report, data as of August 31, 1983. Volume 8, No. 9

    International Nuclear Information System (INIS)

    1984-10-01

    This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  20. Corrosion problem in the CRENK Triga Mark II research reactor

    International Nuclear Information System (INIS)

    Kalenga, M.

    1990-01-01

    In August 1987, a routine underwater optical inspection of the aluminum tank housing the core of the CRENK Triga Mark II reactor, carried out to update safety condition of the reactor, revealed pitting corrosion attacks on the 8 mm thick aluminum tank bottom. The paper discuss the work carried out by the reactor staff to dismantle the reactor in order to allow a more precise investigation of the corrosion problem, to repair the aluminum tank bottom, and to enhance the reactor overall safety condition

  1. Simultaneous hydrogen utilization and in situ biogas upgrading in an anaerobic reactor

    DEFF Research Database (Denmark)

    Luo, Gang; Johansson, Sara; Boe, Kanokwan

    2012-01-01

    . The methane production rate of the reactor with H2 addition was 22% higher, compared to the control reactor only fed with manure. The CO2 content in the produced biogas was only 15%, while it was 38% in the control reactor. However, the addition of hydrogen resulted in increase of pH (from 8.0 to 8.3) due......The possibility of converting hydrogen to methane and simultaneous upgrading of biogas was investigated in both batch tests and fully mixed biogas reactor, simultaneously fed with manure and hydrogen. Batch experiments showed that hydrogen could be converted to methane by hydrogenotrophic...

  2. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  3. Advances in Reactor Physics, Mathematics and Computation. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume one, are divided into 6 sessions bearing on: - session 1: Advances in computational methods including utilization of parallel processing and vectorization (7 conferences) - session 2: Fast, epithermal, reactor physics, calculation, versus measurements (9 conferences) - session 3: New fast and thermal reactor designs (9 conferences) - session 4: Thermal radiation and charged particles transport (7 conferences) - session 5: Super computers (7 conferences) - session 6: Thermal reactor design, validation and operating experience (8 conferences).

  4. Analysis of Kinetic Parameter Effect on Reactor Operation Stability of the RSG-GAS Reactor

    International Nuclear Information System (INIS)

    Rokhmadi

    2007-01-01

    Kinetic parameter has influence to behaviour on RSG-GAS reactor operation. In this paper done is the calculation of reactivity curve, period-reactivity relation and low power transfer function in silicide fuel. This parameters is necessary and useful for reactivity characteristic analysis and reactor stability. To know the reactivity response, it was done reactivity insertion at power 1 watt using POKDYN code because at this level of power no feedback reactivity so important for reactor operation safety. The result of calculation showed that there is no change of significant a period-reactivity relation and transfer function at low power for 2.96 gU/cc, 3.55 gU/cc and 4.8 gU/cc density of silicide fuels. The result of the transfer function at low power showed that the reactor is critical stability with no feedback. The result of calculation also showed that reactivity response no change among three kinds of fuel densities. It can be concluded that from kinetic parameter point of view period-reactivity relation, transfer function at low power, and reactivity response are no change reactor operation from reactivity effect when fuel exchanged. (author)

  5. Research nuclear reactor RA - Annual report 1992

    International Nuclear Information System (INIS)

    Sotic, O.

    1992-12-01

    Research reactor RA Annual report for year 1992 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. Second annex B is a paper by Z. Vukadin 'Recurrence formulas for evaluating expansion series of depletion functions' published in 'Kerntechnik' 56, (1991) No.6 (INIS record no. 23024136. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data [sr

  6. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    Energy Technology Data Exchange (ETDEWEB)

    Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

    1999-08-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

  7. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    International Nuclear Information System (INIS)

    Hastowo, Hudi; Tarigan, Alim

    1999-01-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U 3 O 8 -Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

  8. Improved performance of parallel surface/packed-bed discharge reactor for indoor VOCs decomposition: optimization of the reactor structure

    International Nuclear Information System (INIS)

    Jiang, Nan; Hui, Chun-Xue; Li, Jie; Lu, Na; Shang, Ke-Feng; Wu, Yan; Mizuno, Akira

    2015-01-01

    The purpose of this paper is to develop a high-efficiency air-cleaning system for volatile organic compounds (VOCs) existing in the workshop of a chemical factory. A novel parallel surface/packed-bed discharge (PSPBD) reactor, which utilized a combination of surface discharge (SD) plasma with packed-bed discharge (PBD) plasma, was designed and employed for VOCs removal in a closed vessel. In order to optimize the structure of the PSPBD reactor, the discharge characteristic, benzene removal efficiency, and energy yield were compared for different discharge lengths, quartz tube diameters, shapes of external high-voltage electrode, packed-bed discharge gaps, and packing pellet sizes, respectively. In the circulation test, 52.8% of benzene was removed and the energy yield achieved 0.79 mg kJ −1 after a 210 min discharge treatment in the PSPBD reactor, which was 10.3% and 0.18 mg kJ −1 higher, respectively, than in the SD reactor, 21.8% and 0.34 mg kJ −1 higher, respectively, than in the PBD reactor at 53 J l −1 . The improved performance in benzene removal and energy yield can be attributed to the plasma chemistry effect of the sequential processing in the PSPBD reactor. The VOCs mineralization and organic intermediates generated during discharge treatment were followed by CO x selectivity and FT-IR analyses. The experimental results indicate that the PSPBD plasma process is an effective and energy-efficient approach for VOCs removal in an indoor environment. (paper)

  9. Proceedings of the 1992 topical meeting on advances in reactor physics

    International Nuclear Information System (INIS)

    1992-01-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements ampersand Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  10. Activation product transport in fusion reactors

    International Nuclear Information System (INIS)

    Klein, A.C.; Vogelsang, W.F.

    1984-01-01

    Activated corrosion and neutron sputtering products will enter the coolant and/or tritium breeding material of fusion reactor power plants and experiments and cause personnel access problems. Radiation levels around plant components due to these products will cause difficulties with maintenance and repair operations throughout the plant. A computer code, RAPTOR, has been developed to determine the transport of these products in fusion reactor coolant/tritium breeding materials. Without special treatment, it is likely that fusion reactor power plant operators could experience dose rates as high as 8 rem per hour around a number of plant components after only a few years of operation. (orig.)

  11. Fission product release from SLOWPOKE-2 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

    1994-12-31

    Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

  12. In-reactor deformation and fracture of austenitic stainless steels

    International Nuclear Information System (INIS)

    Bloom, E.E.; Wolfer, W.G.

    1978-01-01

    An experimental technique for determining in-reactor fracture strain was developed and demonstrated. Differential swelling between a sample holder and a test specimen with a lower swelling rate produced uniaxial deformation. In-reactor deformations of 0.7 to 2.1% were achieved in type 304 stainless steel previously irradiated to fluences up to 8.8 x 10 26 n/m 2 without fracture. These strains are significantly higher than found in postirradiation creep-rupture tests on similar samples. From the measured strain values and published irradiation creep data and correlations, the stress levels during the irradiation were calculated. On the basis of previous postirradiation creep-rupture results, many of the samples that did not fail would be predicted to fail. Thus we conclude that the in-reactor rupture life is longer than predicted by postirradiation tests. Strain in a fractured sample was estimated to be less than 3.8%, and the in-reactor fractures were intergranular--the same fracture mode as found in postirradiation tests. Irradiation creep may relax stresses at crack tips and sliding boundaries, thus retarding the initiation and/or growth of cracks and leading to longer rupture lives in-reactor. However, the very high ductility or superplastic behavior predicted by the strain rate sensitivity of irradiation creep is not achieved because of the eventual interruption of the deformation process by grain boundary fracture

  13. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

    In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

  14. Introducing Zenith project

    International Nuclear Information System (INIS)

    Dupejova, E.

    2008-01-01

    Zenith ia a international program of continuous improvement, aimed at reaching high work performance in all ENEL activities and in all countries where ENEL is present Zenith covers various projects aimed at cost reduction, processes simplification and performance improvement that have been already initiated in the past (DTC, TTC, QUASAR...) The program includes 5 pillars to reach perfection: - reaching for operational proficiency by underlining the leading role of ENEL and continuous searching possibilities how to improve processes and activities; - to learn methods of continuous the others; - to learn and apply the methods of improvement; - continuously apply the methods of improvement; - continuously monitor and publish the results and implementation of ZENITH standard. (author)

  15. Core configuration of a gas-cooled reactor as a tritium production device for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakaya, H., E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, H.; Nakao, Y. [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Shimakawa, S.; Goto, M.; Nakagawa, S. [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan); Nishikawa, M. [Malaysia-Japan International Institute of Technology, UTM, Kuala Lumpur 54100 (Malaysia)

    2014-05-01

    The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO{sub 2} as the tritium-producing material. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO{sub 2} is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3 GW can produce almost 8 kg of tritium in a year.

  16. Tokamak engineering test reactor

    International Nuclear Information System (INIS)

    Conn, R.W.; Jassby, D.L.

    1975-07-01

    The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m 2 is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m 2 . The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10 12 cm -3 s. The plasma temperature is maintained by injection of 177 MW of 200-keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10 -6 . If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the []outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m 2 can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth)

  17. Completion of the privatisation

    International Nuclear Information System (INIS)

    Anon

    2006-01-01

    In Bratislava on April 28, 2006 the privatisation of 66 % share of the Slovenske Elektrarne, a. s. (SE) has definitively finished. The Slovak Republic with the Italian company ENEL assigned ultimate documents and the ENEL definitively entered into the SE. Italian investor paid for the SE 839 millions EUR, what is about 31.35 billions Slovak crowns. (author)

  18. Air pollution restrictions in electrical production system

    International Nuclear Information System (INIS)

    Gallizioli, G.

    1993-01-01

    A description of the principal characteristics regarding the Italian electrical power system and the evolution of standardization in air pollution control is given. Afterwards, ENEL (the Italian National Electricity Board) actions in the environmental protection field (with particular respect to thermo-electrical production) are presented. Finally, principal ENEL research programs on new air pollution control technologies are discussed

  19. Licensed operating reactors. Status summary report: data as of July 31, 1985. Volume 9, No. 8

    International Nuclear Information System (INIS)

    1985-09-01

    This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Head-quarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capabilitiy, reactor years of experience and non-power reactors in the United States

  20. Operation and maintenance of 1MW PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    Adnan Bokhari; Mohammad Suhaimi Kassim

    2006-01-01

    The Malaysian Research Reactor, Reactor TRIGA PUSPATI (RTP) has been successfully operated for 22 years for various experiments. Since its commissioning in June 1982 until December 2004, the 1MW pool-type reactor has accumulated more than 21143 hours of operation, corresponding to cumulative thermal energy release of about 14083 MW-hours. The reactor is currently in operation and normally operates on demand, which is normally up to 6 hours a day. Presently the reactor core is made up of standard TRIAGA fuel element consists of 8.5 wt%, 12 wt% and 20 wt% types; 20%-enriched and stainless steel clad. Several measures such as routine preventive maintenance and improving the reactor support systems have been taken toward achieving this long successful operation. Besides normal routine utilization like other TRIGA reactors, new strategies are implemented for effective increase in utilization. (author)

  1. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  2. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

    A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  4. Experience with spent fuel storage at research and test reactors. Proceedings of an advisory group meeting held in Vienna, 5-8 July 1993

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-01-01

    Irradiated fuel from research and test reactors has been stored at various facilities for several decades. As these facilities age and approach or exceed their original design lifetimes, there is mounting concern about closure of the fuel cycle and about the integrity of ageing fuels from the materials point of view as well as some concern about the loss of self-protection of the fuels as their activity decays. It is clear that an international effort is necessary to give these problems sufficient exposure and to ensure that work begins on appropriate solutions. To obtain an overall picture of the size and extent of these problems, an Advisory Group Meeting on Storage Experience with Spent Fuel from Research Reactors was convened in Vienna 5-8 July 1993, and attended by twelve participants and three observers representing thirteen different countries. These proceedings contain the country reports presented at the meeting. Refs, figs and tabs.

  5. Experience with spent fuel storage at research and test reactors. Proceedings of an advisory group meeting held in Vienna, 5-8 July 1993

    International Nuclear Information System (INIS)

    1995-01-01

    Irradiated fuel from research and test reactors has been stored at various facilities for several decades. As these facilities age and approach or exceed their original design lifetimes, there is mounting concern about closure of the fuel cycle and about the integrity of ageing fuels from the materials point of view as well as some concern about the loss of self-protection of the fuels as their activity decays. It is clear that an international effort is necessary to give these problems sufficient exposure and to ensure that work begins on appropriate solutions. To obtain an overall picture of the size and extent of these problems, an Advisory Group Meeting on Storage Experience with Spent Fuel from Research Reactors was convened in Vienna 5-8 July 1993, and attended by twelve participants and three observers representing thirteen different countries. These proceedings contain the country reports presented at the meeting. Refs, figs and tabs

  6. Statistic method of research reactors maximum permissible power calculation

    International Nuclear Information System (INIS)

    Grosheva, N.A.; Kirsanov, G.A.; Konoplev, K.A.; Chmshkyan, D.V.

    1998-01-01

    The technique for calculating maximum permissible power of a research reactor at which the probability of the thermal-process accident does not exceed the specified value, is presented. The statistical method is used for the calculations. It is regarded that the determining function related to the reactor safety is the known function of the reactor power and many statistically independent values which list includes the reactor process parameters, geometrical characteristics of the reactor core and fuel elements, as well as random factors connected with the reactor specific features. Heat flux density or temperature is taken as a limiting factor. The program realization of the method discussed is briefly described. The results of calculating the PIK reactor margin coefficients for different probabilities of the thermal-process accident are considered as an example. It is shown that the probability of an accident with fuel element melting in hot zone is lower than 10 -8 1 per year for the reactor rated power [ru

  7. Gas-cooled reactors: the importance of their development

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1979-06-01

    The nearest term GCR is the steam-cycle HTGR, which can be used for both power and process steam production. Use of SC-HTGRs permits timely introduction of thorium fuel cycles and of high-thermal-efficiency reactors, decreasing the need for mined U 3 O 8 before arrival of symbiotic fueling of fast-thermal reactor systems. The gas-turbine HTGR offers prospects of lower capital costs than other nuclear reactors, but it appears to require longer and more costly development than the SC-HTGR. Accelerated development of the GT-HTGR is needed to gain the advantages of timely introduction. The Gas-Cooled Fast Breeder Reactor (GCFR) offers the possibility of fast breeder reactors with lower capital costs and with higher breeding ratios from oxide fuels. The VHTR provides high-temperature heat for hydrogen production

  8. Grey water treatment in upflow anaerobic sludge blanket (UASB) reactor at different temperatures.

    Science.gov (United States)

    Elmitwalli, Tarek; Otterpohl, Ralf

    2011-01-01

    The treatment of grey water in two upflow anaerobic sludge blanket (UASB) reactors, operated at different hydraulic retention times (HRTs) and temperatures, was investigated. The first reactor (UASB-A) was operated at ambient temperature (14-25 degrees C) and HRT of 20, 12 and 8 h, while the second reactor (UASB-30) was operated at controlled temperature of 30 degrees C and HRT of 16, 10 and 6 h. The two reactors were fed with grey water from 'Flintenbreite' settlement in Luebeck, Germany. When the grey water was treated in the UASB reactor at 30 degrees C, total chemical oxygen demand (CODt) removal of 52-64% was achieved at HRT between 6 and 16 h, while at lower temperature lower removal (31-41%) was obtained at HRT between 8 and 20 h. Total nitrogen and phosphorous removal in the UASB reactors were limited (22-36 and 10-24%, respectively) at all operational conditions. The results showed that at increasing temperature or decreasing HRT of the reactors, maximum specific methanogenic activity of the sludge in the reactors improved. As the UASB reactor showed a significantly higher COD removal (31-64%) than the septic tank (11-14%) even at low temperature, it is recommended to use UASB reactor instead of septic tank (the most common system) for grey water pre-treatment. Based on the achieved results and due to high peak flow factor, a HRT between 8 and 12 h can be considered the suitable HRT for the UASB reactor treating grey water at temperature 20-30 degrees C, while a HRT of 12-24 h can be applied at temperature lower than 20 degrees C.

  9. 8. stellarator workshop

    International Nuclear Information System (INIS)

    1991-07-01

    The technical reports in this collection of papers were presented at the 8th International Workshop on Stellarators, and International Atomic Energy Agency Technical Committee Meeting. They include presentations on transport, magnetic configurations, fluctuations, equilibrium, stability, edge plasma and wall aspects, heating, diagnostics, new concepts and reactor studies. Refs, figs and tabs

  10. Shielding design to obtain compact marine reactor

    International Nuclear Information System (INIS)

    Yamaji, Akio; Sako, Kiyoshi

    1994-01-01

    The marine reactors equipped in previously constructed nuclear ships are in need of the secondary shield which is installed outside the containment vessel. Most of the weight and volume of the reactor plants are occupied by this secondary shield. An advanced marine reactor called MRX (Marine Reactor X) has been designed to obtain a more compact and lightweight marine reactor with enhanced safety. The MRX is a new type of marine reactor which is an integral PWR (The steam generator is installed in the pressure vessel.) with adopting a water-filled containment vessel and a new shielding design method of no installation of the secondary shield. As a result, MRX is considerably lighter in weight and more compact in size as compared with the reactors equipped in previously constructed nuclear ships. For instance, the plant weight and volume of the containment vessel of MRX are about 50% and 70% of those of the Nuclear Ship MUTSU, in spite of the power of MRX is 2.8 times as large as the MUTSU's reactor. The shielding design calculation was made using the ANISN, DOT3.5, QAD-CGGP2 and ORIGEN codes. The computational accuracy was confirmed by experimental analyses. (author)

  11. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  12. Main activities carried out for the conversion of the reactor core TRIGA, from HEU 8.5/70 / LEU 8.5/20 to LEU 30/20; Principales actividades llevadas a cabo para la conversion del nucleo del reactor TRIGA, de HEU 8.5/70 / LEU 8.5/20 a LEU 30/20

    Energy Technology Data Exchange (ETDEWEB)

    Flores C, J., E-mail: jorge.floresc@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    In agreement with the policies of the global initiative of threats reduction (GTRI), Mexico committed that inside the reduction program of the fuel enrichment in research and test reactors (RERTR), the conversion of the core reactor TRIGA (in the nuclear centre) would be made, to use solely fuel with low enrichment ({<=} 20% U{sup 235}). To support to the execution of this commitment, a series of accords and agreements were established. The Project Agreement and Supply among the IAEA, the United States of America and Mexico was the more relevant. In this work the main activities carried out in the Instituto Nacional de Investigaciones Nucleares (ININ) with this purpose are presented. (Author)

  13. Research and development studies on the seismic behaviour of the PEC fast reactor (safety analysis detailed report no. 8)

    Energy Technology Data Exchange (ETDEWEB)

    Martelli, A.; Forni, M.; Masoni, P.; Maresca, G.; Castoldi, A.; Muzzi, F. [ENEA, Rome (Italy); Ansaldo Spa, Genoa [Italy; ISMES Spa, Bergamo [Italy

    1988-01-15

    This paper presents the main features and results of the numerical and experimental studies that were carried out by ENEA (Italian Commission for Alternative Energy Sources) for the seismic verification of the Italian PEC fast reactor test facility. More precisely, the paper focuses on the wide-ranging research and development programme that has been performed (and recently completed) on the reactor building, the reactor-block, the main vessel, the core and the shutdown system. The needs of these detailed studies are stressed and the feed-backs on the design, necessary safisfy the seismic safety requirements, are recalled. The general validity of the analyses in the framework of the research and development activities for nuclear reactor is also pointed out.

  14. Nuclear reactors

    International Nuclear Information System (INIS)

    Yoshioka, Michiko.

    1985-01-01

    Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

  15. Biogas production from UASB and polyurethane carrier reactors treating sisal processing wastewater

    Energy Technology Data Exchange (ETDEWEB)

    Rubindamayugi, M S.T.; Salakana, L K.P. [Univ. of Dar es Salaam, Faculty of Science, Applied Microbiology Unit (Tanzania, United Republic of)

    1998-12-31

    The fundamental benefits which makes anaerobic digestion technology (ADT) attractive to the poor developing include the low cost and energy production potential of the technology. In this study the potential of using UASB reactor and Polyurethane Carrier Reactor (PCR) as pollution control and energy recovery systems from sisal wastewater were investigated in lab-scale reactors. The PCR demonstrated the shortest startup period, whereas the UASB reactor showed the highest COD removal efficiency 79%, biogas production rate (4.5 l biogas/l/day) and process stability than the PCR under similar HRT of 15 hours and OLR of 8.2 g COD/l/day. Both reactor systems became overloaded at HRT of 6 hours and OLR of 15.7 g COD/l/day, biogas production ceased and reactors acidified to pH levels which are inhibiting to methanogenesis. Based on the combined results on reactor performances, the UASB reactor is recommended as the best reactor for high biogas production and treatment efficiency. It was estimated that a large-scale UASB reactor can be designed under the same loading conditions to produce 2.8 m{sup 3} biogas form 1 m{sup 3} of wastewater of 5.16 kg COD/m{sup 3}. Wastewater from one decortication shift can produce 9,446 m{sup 3} og biogas. The energy equivalent of such fuel energy is indicated. (au)

  16. Biogas production from UASB and polyurethane carrier reactors treating sisal processing wastewater

    Energy Technology Data Exchange (ETDEWEB)

    Rubindamayugi, M.S.T.; Salakana, L.K.P. [Univ. of Dar es Salaam, Faculty of Science, Applied Microbiology Unit (Tanzania, United Republic of)

    1997-12-31

    The fundamental benefits which makes anaerobic digestion technology (ADT) attractive to the poor developing include the low cost and energy production potential of the technology. In this study the potential of using UASB reactor and Polyurethane Carrier Reactor (PCR) as pollution control and energy recovery systems from sisal wastewater were investigated in lab-scale reactors. The PCR demonstrated the shortest startup period, whereas the UASB reactor showed the highest COD removal efficiency 79%, biogas production rate (4.5 l biogas/l/day) and process stability than the PCR under similar HRT of 15 hours and OLR of 8.2 g COD/l/day. Both reactor systems became overloaded at HRT of 6 hours and OLR of 15.7 g COD/l/day, biogas production ceased and reactors acidified to pH levels which are inhibiting to methanogenesis. Based on the combined results on reactor performances, the UASB reactor is recommended as the best reactor for high biogas production and treatment efficiency. It was estimated that a large-scale UASB reactor can be designed under the same loading conditions to produce 2.8 m{sup 3} biogas form 1 m{sup 3} of wastewater of 5.16 kg COD/m{sup 3}. Wastewater from one decortication shift can produce 9,446 m{sup 3} og biogas. The energy equivalent of such fuel energy is indicated. (au)

  17. Delayed photoneutrons of the of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Ngo Quang Huy; Ha Van Thong; Vu Hai Long; Ngo Phu Khang; Nguyen Nhi Dien; Pham Van Lam; Huynh Dong Phuong; Luong Ba Vien; Le Vinh Vinh

    1994-01-01

    Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10 5 /10 8 n/cm 2 /sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to (γ,n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is β B e eff =0.49%β eff for a beryllium weight relative to U 235 fuel of m B e/m U = 8.5. This result is acceptable in comparison to those obtained for other Be-U 235 media. (author). 5 refs., 2 figs., 4 tabs

  18. Nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

  19. Economic analysis of nuclear reactors

    International Nuclear Information System (INIS)

    Owen, P.S.; Parker, M.B.; Omberg, R.P.

    1979-05-01

    The report presents several methods for estimating the power costs of nuclear reactors. When based on a consistent set of economic assumptions, total power costs may be useful in comparing reactor alternatives. The principal items contributing to the total power costs of a nuclear power plant are: (1) capital costs, (2) fuel cycle costs, (3) operation and maintenance costs, and (4) income taxes and fixed charges. There is a large variation in capital costs and fuel expenses among different reactor types. For example, the standard once-through LWR has relatively low capital costs; however, the fuel costs may be very high if U 3 O 8 is expensive. In contrast, the FBR has relatively high capital costs but low fuel expenses. Thus, the distribution of expenses varies significantly between these two reactors. In order to compare power costs, expenses and revenues associated with each reactor may be spread over the lifetime of the plant. A single annual cost, often called a levelized cost, may be obtained by the methods described. Levelized power costs may then be used as a basis for economic comparisons. The paper discusses each of the power cost components. An exact expression for total levelized power costs is derived. Approximate techniques of estimating power costs will be presented

  20. The renaissance of Italian nuclear power; La renaissance du nucleaire italien

    Energy Technology Data Exchange (ETDEWEB)

    Bouchter, J.C.; Cassuto, A. [CEA/Ambassade de France a Rome (Italy)

    2010-07-15

    In the fifties Italy was an advanced country in terms of nuclear electricity but as a consequence of the Chernobyl accident Italy changed drastically its energy policy and closed definitely all its nuclear plants. Now in order to be less dependent on energy imports and to reduce its CO{sub 2} emission, Italy has changed its mind and welcomes nuclear power in its future energy mix. The aim is to reach the following contributions for the production of electricity in 2030: 50% from fossil fuels, 25% from renewable energies and 25% from nuclear energy (13.000 MWe) and with a first reactor operating in 2020. The main actors of the renaissance of nuclear power in Italy are: -) ENEL (the second electricity producer in Europe), -) SOGIN, a company that is mainly in charge of the dismantling of nuclear plants, -) ENEA a state agency for the development of new technologies, energy and sustainable development, and -) companies working in the nuclear industry like ANSALDO. Various collaboration agreements have been signed between ENEL and EDF or between ENEA and CEA concerning staff training, nuclear safety or radioactive waste management. The main difficulties of this renaissance of the nuclear energy are to get the agreement of the national and local populations as well as that of the political class that is strongly marked by a division in 2 wings. (A.C.)

  1. The renaissance of Italian nuclear power

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Cassuto, A.

    2010-01-01

    In the fifties Italy was an advanced country in terms of nuclear electricity but as a consequence of the Chernobyl accident Italy changed drastically its energy policy and closed definitely all its nuclear plants. Now in order to be less dependent on energy imports and to reduce its CO 2 emission, Italy has changed its mind and welcomes nuclear power in its future energy mix. The aim is to reach the following contributions for the production of electricity in 2030: 50% from fossil fuels, 25% from renewable energies and 25% from nuclear energy (13.000 MWe) and with a first reactor operating in 2020. The main actors of the renaissance of nuclear power in Italy are: -) ENEL (the second electricity producer in Europe), -) SOGIN, a company that is mainly in charge of the dismantling of nuclear plants, -) ENEA a state agency for the development of new technologies, energy and sustainable development, and -) companies working in the nuclear industry like ANSALDO. Various collaboration agreements have been signed between ENEL and EDF or between ENEA and CEA concerning staff training, nuclear safety or radioactive waste management. The main difficulties of this renaissance of the nuclear energy are to get the agreement of the national and local populations as well as that of the political class that is strongly marked by a division in 2 wings. (A.C.)

  2. ANAEROBIC DIGESTION AND THE DENITRIFICATION IN UASB REACTOR

    Directory of Open Access Journals (Sweden)

    José Tavares de Sousa

    2008-01-01

    Full Text Available The environmental conditions in Brazil have been contributing to the development of anaerobic systems in the treatment of wastewaters, especially UASB - Upflow Anaerobic Sludge Blanket reactors. The classic biological process for removal of nutrients uses three reactors - Bardenpho System, therefore, this work intends an alternative system, where the anaerobic digestion and the denitrification happen in the same reactor reducing the number of reactors for two. The experimental system was constituted by two units: first one was a nitrification reactor with 35 L volume and 15 d of sludge age. This system was fed with raw sanitary waste. Second unit was an UASB, with 7.8 L and 6 h of hydraulic detention time, fed with ¾ of effluent nitrification reactor and ¼ of raw sanitary waste. This work had as objective to evaluate the performance of the UASB reactor. In terms of removal efficiency, of bath COD and nitrogen, it was verified that the anaerobic digestion process was not affected. The removal efficiency of organic material expressed in COD was 71%, performance already expected for a reactor of this type. It was also observed that the denitrification process happened; the removal nitrate efficiency was 90%. Therefore, the denitrification process in reactor UASB is viable.

  3. Use of plate fuel elements for the RA3 reactor

    International Nuclear Information System (INIS)

    Parodi, C.; Parkanski, D.; Higa, M.; Marajofsky, A.

    1992-01-01

    The RA3 reactor is a pool reactor, redesigned for 5 MW dissipation. Nineteen plates are used in each fuel element. The utilization of 20% enriched U, gives the possibility of the development of rod type fuel with Al/U 3 O 8 cermets. The thermohydraulic and neutronic conditions are studied in this work in order to satisfy the stipulated power. In addition, the fabrication conditions of Al/U 3 O 8 and Al/U 3 O 8 /Zr H 2 cermets with densities within the limits imposed by the thermohydraulics and neutronics conditions are studied. (author)

  4. A Compact Quasi-axisymmetric Stellarator Reactor

    International Nuclear Information System (INIS)

    Ku, L.P.

    2003-01-01

    We report the progress made in assessing the potential of compact, quasi-axisymmetric stellarators as power-producing reactors. Using an aspect ratio A=4.5 configuration derived from NCSX and optimized with respect to the quasi-axisymmetry and MHD stability in the linear regime as an example, we show that a reactor of 1 GW(e) maybe realizable with a major radius *8 m. This is significantly smaller than the designs of stellarator reactors attempted before. We further show the design of modular coils and discuss the optimization of coil aspect ratios in order to accommodate the blanket for tritium breeding and radiation shielding for coil protection. In addition, we discuss the effects of coil aspect ratio on the peak magnetic field in the coils

  5. Modular Stellarator Fusion Reactor (MSR) concept

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.

    1981-01-01

    A preliminary conceptual study has been made of the Modulator Stellarator Reactor (MSR) as a stedy-state, ignited, DT-fueled, magnetic fusion reactor. The MSR concept combines the physics of classic stellarator confinement with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4.8-GWt plant based on Alcator transport scaling and an average beta value of 0.04 in an l = 2 system with a plasma aspect ratio of 11. Neither an economic analysis nor a detailed conceptual engineering design is presented here, as the primary intent of this scoping study is the elucidation of key physics tradeoffs, constraints, and uncertainties for the ultimate power-reactor embodiment

  6. Calculation of fundamental parameters for the dynamical study of TRIGA-3-Salazar reactor (Mixed reactor core)

    International Nuclear Information System (INIS)

    Viais J, J.

    1994-01-01

    Kinetic parameters for dynamic study of two different configurations, 8 and 9, both with standard fuel, 20% enrichment and Flip (Fuel Life Improvement Program with 70% enrichment) fuel, for TRIGA Mark-III reactor from Mexico Nuclear Center, are obtained. A calculation method using both WIMS-D4 and DTF-IV and DAC1 was established, to decide which of those two configurations has the best safety and operational conditions. Validation of this methodology is done by calculate those parameters for a reactor core with new standard fuel. Configuration 9 is recommended to be use. (Author)

  7. Fast Reactor Physics. Vol. II. Proceedings of a Symposium on Fast Reactor Physics and Related Safety Problems

    International Nuclear Information System (INIS)

    1968-01-01

    Proceedings of a Symposium organized by the IAEA and held in Karlsruhe, 30 October - 3 November 1967. The meeting was attended by 183 scientists from 23 countries and three international organizations. Contents: (Vol.1) Review of national programmes (5 papers); Nuclear data for fast reactors (12 papers); Experimental methods (3 papers); Zoned systems (7 papers); Kinetics (7 papers). (Vol.11) Fast critical experiments (8 papers); Heterogeneity in fast critical experiments (5 papers); Fast power reactors (13 papers); Fast pulsed reactors (3 papers); Panel discussion. Each paper is in its original language (50 English, 11 French and 3 Russian) 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)

  8. Safety of RBMK reactors: Major results and prospects

    International Nuclear Information System (INIS)

    Sidorenko, V.A.

    1996-01-01

    The paper considers the following issues: basic reasons for the advent of NPPs with RBMK reactors; the logic of identifying top-priority measures immediately after the accident; top-priority measures for improving the safety and reliability of NPPs with RBMK reactors; upgrading NPPs with RBMK reactors in compliance with the Norms; programmes for retrofitting and upgrading of NPPs of the ''Rosnergoatom'' Concern and progress with their implementation as of April 1996; the safety of RBMK plants and the programmes of its enhancement with regard to modern requirements in the light of national and international assessment; objective indicators of safety, reliability, and economic efficiency of NPPs with RBMK reactors; economics: rationale for continuing plants operation till the end of their design lifetime. 8 refs, 3 figs

  9. Delayed photoneutrons of the of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Khang, Ngo Phu; Dien, Nguyen Nhi; Lam, Pham Van; Phuong, Huynh Dong; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10{sup 5}/10{sup 8} n/cm{sup 2}/sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to ({gamma},n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is {beta}{sup B}e{sub eff}=0.49%{beta}{sub eff} for a beryllium weight relative to U{sup 235} fuel of m{sub B}e/m{sub U} = 8.5. This result is acceptable in comparison to those obtained for other Be-U{sup 235} media. (author). 5 refs., 2 figs., 4 tabs.

  10. Photocatalytic reactors for treating water pollution with solar illumination: a simplified analysis for n-steps flow reactors with recirculation

    Energy Technology Data Exchange (ETDEWEB)

    Sagawe, G.; Bahnemann, D. [Universitaet Hannover (Germany). Institut fuer Technische Chemie; Brandi, R.J.; Cassano, A.E. [INTEC Universidad Nacional del Litoral and CONICET, Sante Fe (Argentina)

    2005-09-01

    The concentration of dissolved oxygen in water, in equilibrium with atmospheric air (ca. 8 ppm at 20{sup o}C), defines the limits of all practical oxidizing processes for removing pollutants in photocatalytic reactors. To solve this limitation, an alternative approach to that of a continuously aerated reactor is the use of a recirculating system with aeration performed after every cycle at the reactor entering stream. As defined by the nature of a single recirculating step (the need of a reactor operation at a rather low concentration range), this procedure results in a very low photonic efficiency (thus requiring a large photon collecting area and consequently increasing the capital cost). The design engineer will have to resort to a series of several reactors with recirculation. This solution may then lead to a very high Photonic Efficiency for the entire process (i.e., a reduced light harvesting area) at the price of an increase in the required capital cost (due to the larger number of reactors). This paper provides a very simple analysis and analytical expressions that can be used to estimate, for a desired degree of degradation, a trade-off solution between a high number of reactors and a very large surface area to collect the solar photons. (author)

  11. Quality assurance in the manufacture of metallic uranium fuel for research reactors

    International Nuclear Information System (INIS)

    Shah, B.K.; Kumar, Arbind; Nanekar, P.P.; Vaidya, P.R.

    2009-01-01

    Two Research Reactors viz. CIRUS and DHRUVA are operating at Trombay since 1960 and 1985 respectively. Cirus is a 40 MWth reactor using heavy water as moderator and light water as coolant. Dhruva is a 100 MWth reactor using heavy water as moderator and coolant. The maximum neutron flux of these reactors are 6.7 x 10 13 n/cm 2 /s (Cirus) and 1.8 x 10 14 n/cm 2 /s (Dhruva). Both these reactors are used for basic research, R and D in reactor technology, isotope production and operator training. Fuel material for these reactors is natural uranium metallic rods claded in finned aluminium (99.5%) tubes. This presentation will discuss various issues related to fabrication quality assurance and reactor behavior of metallic uranium fuel used in research reactors

  12. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

    Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

    2003-01-01

    Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

  13. Catalytic-Dielectric Barrier Discharge Plasma Reactor For Methane and Carbon Dioxide Conversion

    Directory of Open Access Journals (Sweden)

    Istadi Istadi

    2007-10-01

    Full Text Available A catalytic - DBD plasma reactor was designed and developed for co-generation of synthesis gas and C2+ hydrocarbons from methane. A hybrid Artificial Neural Network - Genetic Algorithm (ANN-GA was developed to model, simulate and optimize the reactor. Effects of CH4/CO2 feed ratio, total feed flow rate, discharge voltage and reactor wall temperature on the performance of catalytic DBD plasma reactor was explored. The Pareto optimal solutions and corresponding optimal operating parameters ranges based on multi-objectives can be suggested for catalytic DBD plasma reactor owing to two cases, i.e. simultaneous maximization of CH4 conversion and C2+ selectivity, and H2 selectivity and H2/CO ratio. It can be concluded that the hybrid catalytic DBD plasma reactor is potential for co-generation of synthesis gas and higher hydrocarbons from methane and carbon dioxide and showed better than the conventional fixed bed reactor with respect to CH4 conversion, C2+ yield and H2 selectivity for CO2 OCM process. © 2007 BCREC UNDIP. All rights reserved.[Presented at Symposium and Congress of MKICS 2007, 18-19 April 2007, Semarang, Indonesia][How to Cite: I. Istadi, N.A.S. Amin. (2007. Catalytic-Dielectric Barrier Discharge Plasma Reactor For Methane and Carbon Dioxide Conversion. Bulletin of Chemical Reaction Engineering and Catalysis, 2 (2-3: 37-44.  doi:10.9767/bcrec.2.2-3.8.37-44][How to Link/DOI: http://dx.doi.org/10.9767/bcrec.2.2-3.8.37-44 || or local: http://ejournal.undip.ac.id/index.php/bcrec/article/view/8][Cited by: Scopus 1 |

  14. U.S. Status of Fast Reactor Research and Technology

    International Nuclear Information System (INIS)

    Hill, Robert

    2012-01-01

    Summary: • Fast reactor R&D is focused on key technologies innovations for performance improvement (cost reduction) and safety: 1. System Integration and Concept Development; 2. Safety Technology; 3. Advanced Materials; 4. Ultrasonic Viewing; 5. Advanced Energy Conversion (Supercritical CO 2 Brayton cycle); 6. Reactor Simulation; 7. Nuclear Data; 8. Advanced Fuels. • Fast reactors have flexible capability for actinide management: – A wide variety of fuel cycle options are being considered; • International R&D collaboration pursued in Generation-IV and multilateral arrangements

  15. The plutonium recycle for PWR reactors from brazilian nuclear program

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-01-01

    The purpose of this thesis is to evaluate the material requirements of the nuclear fuel cycle with plutonium recycle. The study starts with the calculation of a reference reactor and has flexibility to evaluate the demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): Without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5% U 3 O 8 and 6% separative work units if recycle is assumed only after the fifth operation cycle of the thermal reactors. (author)

  16. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  17. Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

    International Nuclear Information System (INIS)

    Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

    2009-01-01

    A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m 3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m 3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m 3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m 3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m 3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m 3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

  18. Numerical modeling of a nuclear production reactor cooling lake

    International Nuclear Information System (INIS)

    Hamm, L.L.; Pepper, D.W.

    1987-01-01

    A finite element model has been developed which predicts flow and temperature distributions within a nuclear reactor cooling lake at the Savannah River Plant near Aiken, South Carolina. Numerical results agree with values obtained from a 3-D EPA numerical lake model and actual measurements obtained from the lake. Because the effluent water from the reactor heat exchangers discharges directly into the lake, downstream temperatures at mid-lake could exceed the South Carolina DHEC guidelines for thermal exchanges during the summer months. Therefore, reactor power was reduced to maintain temperature compliance at mid-lake. Thermal mitigation measures were studied that included placing a 6.1 m deep fabric curtain across mid-lake and moving the reactor outfall upstream. These measurements were calculated to permit about an 8% improvement in reactor power during summer operation

  19. Protective actions as a factor in power reactor siting

    Energy Technology Data Exchange (ETDEWEB)

    Gant, K.S.; Schweitzer, M.

    1984-06-01

    This report examines the relationship between a power reactor site and the ease of implementing protective actions (emergency measures a serious accident). Limiting populating density around a reactor lowers the number of people at risk but cannot assure that all protective actions are possible for those who reside near the reactor. While some protective measures can always be taken (i.e., expedient respiratory protection, sheltering) the ability to evacuate the area or find adequate shelter may depend on the characteristics of the area near the reactor site. Generic siting restrictions designed to identify and eliminate these site-specific constraints would be difficult to formulate. The authors suggest identifying possible impediments to protective actions at a proposed reactor site and addressing these problems in the emergency plans. 66 references, 6 figures, 8 tables.

  20. Protective actions as a factor in power reactor siting

    International Nuclear Information System (INIS)

    Gant, K.S.; Schweitzer, M.

    1984-06-01

    This report examines the relationship between a power reactor site and the ease of implementing protective actions (emergency measures a serious accident). Limiting populating density around a reactor lowers the number of people at risk but cannot assure that all protective actions are possible for those who reside near the reactor. While some protective measures can always be taken (i.e., expedient respiratory protection, sheltering) the ability to evacuate the area or find adequate shelter may depend on the characteristics of the area near the reactor site. Generic siting restrictions designed to identify and eliminate these site-specific constraints would be difficult to formulate. The authors suggest identifying possible impediments to protective actions at a proposed reactor site and addressing these problems in the emergency plans. 66 references, 6 figures, 8 tables

  1. Zirconium-hydride solid zero power reactor and its application research

    International Nuclear Information System (INIS)

    Lin Shenghuo; Luo Zhanglin; Su Zhuting

    1994-10-01

    The Zirconium Hydride Solid Zero Power Reactor built at China Institute of Atomic Energy is introduced. In the reactor Zirconium-hydride is used as moderator, plexiglass as reflector and U 3 O 8 with enrichment of 20% as the fuel, Since its initial criticality, the physical characteristics and safety features have been measured with the result showing that the reactor has sound stability and high sensitivity, etc. It has been successfully used for the personnel training and for the testing of reactor control instruments and experiment devices. It also presents the special advantage for the pre-research of some applications

  2. Tritium resources available for fusion reactors

    Science.gov (United States)

    Kovari, M.; Coleman, M.; Cristescu, I.; Smith, R.

    2018-02-01

    The tritium required for ITER will be supplied from the CANDU production in Ontario, but while Ontario may be able to supply 8 kg for a DEMO fusion reactor in the mid-2050s, it will not be able to provide 10 kg at any realistic starting time. The tritium required to start DEMO will depend on advances in plasma fuelling efficiency, burnup fraction, and tritium processing technology. It is in theory possible to start up a fusion reactor with little or no tritium, but at an estimated cost of 2 billion per kilogram of tritium saved, it is not economically sensible. Some heavy water reactor tritium production scenarios with varying degrees of optimism are presented, with the assumption that only Canada, the Republic of Korea, and Romania make tritium available to the fusion community. Results for the tritium available for DEMO in 2055 range from zero to 30 kg. CANDU and similar heavy water reactors could in theory generate additional tritium in a number of ways: (a) adjuster rods containing lithium could be used, giving 0.13 kg per year per reactor; (b) a fuel bundle with a burnable absorber has been designed for CANDU reactors, which might be adapted for tritium production; (c) tritium production could be increased by 0.05 kg per year per reactor by doping the moderator with lithium-6. If a fusion reactor is started up around 2055, governments in Canada, Argentina, China, India, South Korea and Romania will have the opportunity in the years leading up to that to take appropriate steps: (a) build, refurbish or upgrade tritium extraction facilities; (b) extend the lives of heavy water reactors, or build new ones; (c) reduce tritium sales; (d) boost tritium production in the remaining heavy water reactors. All of the alternative production methods considered have serious economic and regulatory drawbacks, and the risk of diversion of tritium or lithium-6 would also be a major concern. There are likely to be serious problems with supplying tritium for future

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  4. The CAREM reactor and present currents in reactor design

    International Nuclear Information System (INIS)

    Ordonez, J.P.

    1990-01-01

    INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

  5. Ten year report on the operation of the Latina nuclear power station 1964-1973

    International Nuclear Information System (INIS)

    1976-07-01

    The final report on the operation of the Latina nuclear power station, required under the terms of the contract of participation between ENEL and EURATOM is presented. It covers the first ten years of commercial operation (1 January 1964-31 December 1973) of this power station. Latina uses a British Magnox-type gas-graphite natural uranium reactor with a design thermal capacity of 724 MW. The rated electrical output of the three main turbogenerators was originally 210 MW (3x70), but was reduced to 160 MW in 1971. Construction began in November 1958 and was completed when the reactor first reached criticality in December 1962, the station being connected to the Italian electricity network for the first time in May 1963. The gross rated output of 210 MWe was reached in December 1963 and commercial operation began on 1 January 1964, by which date, however, the power station had already fed 295.5 million kWh into the network

  6. Calculations concerning the capability of passive recombiners to control hydrogen concentration in the containment of an advanced PWR

    International Nuclear Information System (INIS)

    Fineschi, F.; Vanini, P.

    1997-01-01

    The Department of Mechanical and Nuclear Constructions of the University of Pisa has developed a computer code, HOCRA, which is able to make an initial evaluation of the capability of catalytic recombiners to remove hydrogen from the atmosphere of the safety containments of nuclear reactors in accident conditions. The code allows the analysis of the average concentration transient of hydrogen in a generic compartment of a safety containment in a nuclear reactor. The software is structured into two groups. The first, mode-1, analyses the average concentration in all the free volume of the containment before a possible venting., whereas the second, mode-2, analyses the average concentration transient in a containment compartment, assuming input and output flow rates into and from the compartment itself The first part of this paper outlines the physical and mathematical model of the code, the second part reports calculations made for an advanced PWR in cooperation with ENEL. (author)

  7. Research reactors for the social safety and prosperous neutron use

    International Nuclear Information System (INIS)

    Ito, Yasuo

    2000-01-01

    The present status of nuclear reactors in Japan and the world was briefly described in this report. Aiming to construct a background of stable future society dependent on nuclear energy, the necessity to establish an organization for research reactors in Japan was pointed out. There are a total of 468 reactors in the world, but only 248 of them are running at present and most of them are superannuated. In Japan, 15 research reactors are running and 8 of them are under collaborative utilization, but not a few of them have various problems. In the education of atomic energy, a reactor is dispensable for understanding its working principle through practice learning. Furthermore, a research reactor has important roles for development of power reactor in addition to various basic studies such as activation analysis, fission track, biological irradiation, neutron scattering, etc. Application of a reactor has been also progressing in industrial and medical fields. However, operation of the reactors has become more and more difficult in Japan because of a large running cost and a lack of residential consensus for nuclear reactor. Here, the author proposed an establishment of organization of research reactor in order to promote utilization of a reactor in the field of education, rearing of professionals and science and engineering. (M.N.)

  8. Towards EPR (European pressurized reactor)

    International Nuclear Information System (INIS)

    Anon.

    2003-01-01

    According to the French industry minister, it is nonsense continuing delaying the construction of an EPR prototype because France needs it in order to renew timely its park of nuclear reactors. The renewing is expected to begin in 2020 and will be assured with third generation reactors like EPR. A quick launching of the EPR prototype is necessary to have it being in service by 2012, the feedback operating experience that will be accumulated over the 8 years that will follow will be necessary to optimize the industrial version and to have it ready by 2020. The EPR reactor has indisputable assets: modern, safer, more competitive and it will produce less wastes than present nuclear reactors. The construction cost of an EPR prototype is estimated to 3 milliard Euros and the nuclear industry operators propose to finance it completely. The EPR prototype does not jeopardize the ambitious French program about renewable energy sources, France is committed to produce 21% of its electricity from renewable energies by 2010 and 10 milliard Euros will be invested over this period on wind energy. Nuclear energy and alternative energies must be considered as 2 aspects of a diversified energy policy. (A.C.)

  9. Neutron behavior, reactor control, and reactor heat transfer. Volume four

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

  10. Reactor science and technology: operation and control of reactors

    International Nuclear Information System (INIS)

    Qiu Junlong

    1994-01-01

    This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

  11. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  12. Report of blind start-up experiments carried out on the reactor Cabri between 4. and 8. July 1966

    International Nuclear Information System (INIS)

    Filipczak, N.; Filipczak, W.; Furet, J.; Kaiser, J.

    1967-01-01

    The blind start-up of a reactor without any neutronic data concerning a relatively wide range of power dynamics can be necessary when difficulties arise in the positioning of the detector or in neutron-gamma discrimination near the multiplying medium. The object of the experiments carried out on the reactor Cabri was to check the very complete analysis of the start-up accident which was studied on an analogue computer. The number of experiments carried out (12) is not sufficient to allow a definite conclusion. Nevertheless the blind start-up method advocated by N. FILIPCZAK and W. FILIPCZAK does not appear to be incompatible with the security during the operational phase (on condition that its dynamic characteristics are close to that of the reactor Cabri). (authors) [fr

  13. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  14. Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

    International Nuclear Information System (INIS)

    Hirayama, Shoichi; Kawabe, Takaya

    1995-01-01

    The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

  15. Reactor physics aspects of CANDU reactors

    International Nuclear Information System (INIS)

    Critoph, E.

    1980-01-01

    These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

  16. Caramel fuel for research reactors

    International Nuclear Information System (INIS)

    Bussy, P.

    1979-11-01

    This fuel for research reactors is made of UO 2 pellets in a zircaloy cladding to replace 93% enriched uranium. It is a cold fuel, non contaminating and non proliferating, enrichment is only 7 to 8%. Irradiation tests were performed until burn-up of 50000 MWD/t [fr

  17. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  18. Electric Energy Consumption of Multi Purpose Reactor GA. Siwabessy During Reactor Operation

    International Nuclear Information System (INIS)

    Koes Indrakoesoema

    2012-01-01

    Electrical power supply of Reactor Center Multi Purpose obtained from PT PLN to 3030 kVA power contracts. Distribution to existing loads in PRSG divided into 3 (three) lines, each of which is supplied through a transformer BHT01, BHT02 and BHT03, each transformer have capacity of 1600 kVA. During reactor operation, only 2 lines that serve loads, each line serve 2 primary pump motor and 2 secondary pump motor. Electrical power for 24 hours for measurement BHT01, the average is 288 kW, for BHT02 is 641 kW and BHT03 is 466 kW. The energy absorbed by each transformer for 24 hours of measurement, for BHT01 is 6.44 MWh, BHT02 absorb 14.8 MWh and BHT03 absorb 10.9 MWh. (author)

  19. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

  20. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  1. Damage by radiation in structural materials of BWR reactor vessels; Dano por radiacion en materiales estructurales de vasijas de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2002-07-01

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA Mark III Salazar reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A{sup 2}). (Author)

  2. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    Kouts, H.

    1986-01-01

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

  3. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  4. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  5. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  6. The status of the PIK reactor

    Energy Technology Data Exchange (ETDEWEB)

    Petrov, Yu V [Academy of Sciences of Russia, Petersburg Nuclear Physics Institute, Gatchina, St. Petersburg (Russian Federation)

    1992-07-01

    This report describes the 100 MW research reactor PIK which is now under construction. The thermal neutron flux in the heavy water reflector exceeds 10{sup 15} cm{sup -2}s{sup -1}; in the light water trap, it is about 4{center_dot}10{sup 15} cm{sup -2}s{sup -1}. The replaceable core vessel allows to vary the parameters of the core over a wide range. The reactor provides sources of hot, cold and ultracold neutrons for 10 horizontal, 6 inclined neutron beams, and 8 neutron guides. At the ends of the beam tubes, the neutron flux is 10{sup 10} - 10{sup 11} cm{sup -2}s{sup -1}. The flux of the long wave neutrons exceeds 10{sup 9} cm{sup -2}s{sup -1}. To ensure precise measurements, the experimental hall is protected against vibrations. The project meets all modern safety requirements. The calculated parameters of the reactor were verified using a full-scale mock-up. Seventy percent of the reactor construction and installation were completed in the beginning of 1992. (author)

  7. Preliminary study or RSG-GAS reactor fuel element integrity

    International Nuclear Information System (INIS)

    Soejoedi, A.; Tarigan, A.; Sujalmo; Prayoga, S.; Suhadi

    1996-01-01

    After 8 years of operation, RSG-GAS was able to reach 15 cycles of reactor operation with 116 irradiated fuels, whereas 49 fuels were produced by NUKEM; and the other 67 were produced by PEBN-BATAN. At the 15 T h cycles, it have been used 40 standard fuels and 8 control fuels (Forty standard fuels and eight control fuels have been used in the 15 t h core cycles). Several activities have been performed in the reactor, to investigate the fuel integrity, among of them are: .fuel visual test with under water camera, which the results were recorder in the video cassette, primary water quality test during, reactor operation, fuel failure detector system examination and compared the PIE results in the Radiometallurgy Installation (RMI). The results showed that the fuel integrity, before and after irradiation, have still good performance and the fission products have not been released yet

  8. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  9. RIA Analysis of Unprotected TRIGA Reactor

    Directory of Open Access Journals (Sweden)

    M.H. Altaf

    2017-07-01

    Full Text Available An RIA (reactivity initiated accident analysis has been carried out for the TRIGA Mark II research reactor considering both step and ramp reactivity ranges within 0.5 % dk/k (< $1 to 2.0 % dk/k (>$2. The insertion time was set at 10 s. Based on the fact that a reactor becomes unprotected if scram does not work at the event of danger, to define unprotected conditions, the time to actuate scram (trip was taken as close to total simulation time. In this long duration of scram inactivity, it is obtained from the present analysis that the reactor remained safe to up to 1.8 % dk/k ($2.57 for step reactivity and 1.99 % dk/k ($2.84 for ramp reactivity. In addition to negative temperature coefficient of reativity, probably the longer time of reactivity insertion keeps TRIGA safe even at larger magnitudes of reactivity during unprotected reactor transients. Coupled point kinetics, neutronics, and thermal hydraulics code EUREKA-2/R has been utilized for this work. It appears that EUREKA-2/RR predicts the sequence of unprotected transient scenario of TRIGA core with good approximation and the results will definitely be helpful for the reactor operators.

  10. Terrestrial Energy bets on molten salt reactors

    International Nuclear Information System (INIS)

    Anon.

    2015-01-01

    Terrestrial Energy is a Canadian enterprise, founded in 2013, for marketing the integral molten salt reactor (IMSR). A first prototype (called MSRE and with an energy output of 8 MW) was designed and operated between 1965 and 1969 by the Oak Ridge National Laboratory. IMSR is a small, modular reactor with a thermal energy output of 400 MW. According to Terrestrial Energy the technology of conventional power reactors is too complicated and too expensive. On the contrary IMSR's technology appears to be simple, easy to operate and affordable. With a staff of 30 people Terrestrial Energy appears to be a start-up in the nuclear sector. A process of pre-licensing will be launched in 2016 with the Canadian nuclear safety authority. (A.C.)

  11. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  12. Civilian Power Program. Part 1, Summary, Current status of reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given

    1959-09-01

    This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

  13. Research reactor standards and their impact on the TRIGA reactor community

    International Nuclear Information System (INIS)

    Richards, W.J.

    1980-01-01

    The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

  14. The influence of the operating schedule of the Greek Research Reactor on the radiological consequences of the reactor

    International Nuclear Information System (INIS)

    Kollas, J.G.

    1986-04-01

    The sensitivity of the radiological consequences of the Greek Research Reactor to the operating schedule of the reactor is assessed in this report. The consequences are due to the occurrence of a postulated accident, a 20% core melt loss of coolant accident. Three different operating schedules are considered: (a) the present 8 hrs/day, 5 days/wk schedule, (b) a 16 hrs/day, 5 days/wk schedule, and (c) a continuous operation schedule. The results of the analysis indicate that there is a direct relation between consequences and duration of operation. (author)

  15. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  16. Estnisk poesi võgar ta plats / Michel Ekman

    Index Scriptorium Estoniae

    Ekman, Michel

    2008-01-01

    Arvustus: Rooste, Jürgen. Presidenten bor nõnstans i närheten / översättning: Peeter Puide. Stockholm : Ellerströms, 2007 ; fs. 2007 / översättning: Peeter Puide. Stockholm : Ellerströms, 2007 ; Kruusa, Kalju. Mötesvis / översättning: Peeter Puide. Stockholm : Ellerströms, 2007 ; Estniska ballader / översättning: Ilmar Laaban. Stockholm : Ellerströms, 2007 ; Laaban, Ilmar. Palingarderomb. Stockholm : Ellerströms, 2007 ; Under, Marie. Gnistor i askan / översättning: Enel Melberg. Stockholm : Ellerströms, 2007 ; Wuolijoki, Hella. Sõngen om kriget / översättning: Enel Melberg. Stockholm : Ellerströms, 2007 ; Alver, Betti. Historien om den vita krõkan / översättning: Enel Melberg. Stockholm : Ellerströms, 2007 ; Kareva, Doris. Tidens gestalt / översättning: Peeter Puide. Stockholm : Ellerströms, 2007

  17. The legal position of the operator and the constructor as regards construction of a nuclear power plant

    International Nuclear Information System (INIS)

    Sartorelli, Claudio

    1976-01-01

    Implementation of the Italian nuclear power programme requires that ENEL (the Italian State Electricity Agency) should solve a series of problems which are not only political, technical and financial, but also legal. The latter include the different positions held by constructors and operators in the nuclear law field. Although the operator must necessarily entrust construction work by contract to specialized agencies, the constructor as a person proper is completely ignored by the legislation in force, while the operator, as the holder of the construction licence is responsible with regard to all the provisions concerning controls, tests, inspections, etc. In recent years ENEL has frequently concluded ''turnkey'' contracts, but in future, given the desirability of diversifying the orders, ENEL may adopt a different policy, and negotiate different contracts with the suppliers of various components. (N.E.A.)

  18. Nuclear fuels for material test reactors

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

    1982-01-01

    Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

  19. Improved nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

  20. Reactor as furnace and reactor as lamp

    International Nuclear Information System (INIS)

    Goldanskii, V.I.

    1992-01-01

    There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

  1. Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Nufarhana Ayuni Joha; Mohd Sabri Minhat

    2012-01-01

    The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

  2. Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

  3. Methods for reactor physics calculations for control rods in fast reactors

    International Nuclear Information System (INIS)

    Grimstone, M.J.; Rowlands, J.L.

    1988-12-01

    The IAEA Specialists' Meeting on ''Methods for Reactor Physics Calculations for Control Rods in Fast Reactors'' was held in Winfrith, United Kingdom, on 6-8 December, 1988. The meeting was attended by 23 participants from nine countries. The purpose of the meeting was to review the current calculational methods and their accuracy as assessed by theoretical studies and comparisons with measurements, and then to identify the requirements for improved methods or additional studies and comparisons. The control rod properties or effects to be considered were their reactivity worths, their effect on the power distribution through the core, and the reaction rates and energy deposition both within and adjacent to the rods. The meeting was divided into five sessions, in the first of which each national delegation presented a brief overview of their programme of work on calculational methods for fast reactor control rods. In the next three sessions a total of seventeen papers were presented describing calculational methods and assessments of their accuracy. The final session was a discussion to draw conclusions regarding the current status of methods and the further developments and validation work required. A separate abstract was prepared for each of the 23 papers presented at the meeting. Refs, figs and tabs

  4. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  5. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  6. Development of a Low Temperature Irradiation Capsule for Research Reactor Materials

    International Nuclear Information System (INIS)

    Choo, Kee Nam; Cho, Man Soon; Lee, Cheol Yong; Yang, Sung Woo; Shin, Yoon Taek; Park, Seng Jae; Kang, Suk Hoon; Kang, Young Hwan; Park, Sang Jun

    2013-01-01

    A new capsule design was prepared and tested at HANARO for a neutron irradiation of core materials of research reactors as a part of the research reactor development project. Irradiation testing of the materials including graphite, beryllium, and zircaloy-4 that are supposed to be used as core materials in research reactors was required for irradiation at up to 8 reactor operation cycles at low temperature (<100 .deg. C). Therefore, three instrumented capsules were designed and fabricated for an evaluation of the neutron irradiation properties of the core materials (Graphite, Be, Zircaloy-4) of research reactors. The capsules were first designed and fabricated to irradiate materials at low temperature (<100 .deg. C) for a long cycle of 8 irradiation cycles at HANARO. Therefore, the safety of the new designed capsule should be fully checked before irradiation testing. Out-pile performance and endurance testing before HANARO irradiation testing was performed using a capsule under a 110% condition of a reactor coolant flow amount. The structural integrity of the capsule was analyzed in terms of a vibration-induced fatigue cracking of a rod tip of the capsule that is suspected to be the most vulnerable part of a capsule. Another two capsules were irradiated at HANARO for 4 cycles, and one capsule was transferred to a hot cell to examine the integrity of the rod tip of the capsule. After confirming the soundness of the 4 cycle-irradiated capsule, the remaining capsule was irradiated at up to 8 cycles at HANARO. Based on the structural integrity analysis of the capsule, an improved capsule design will be suggested for a longer irradiation test at HANARO

  7. Steam coal markets forum - the buyers` outlook

    Energy Technology Data Exchange (ETDEWEB)

    Graybeal, W.R.; Costa, D.; Iyer, S.; Perego, G. [Apex Resources Group (United States)

    1996-12-31

    The session contains four presentations: a talk on the uncertain future of coal, by Bill Graybeal of Apex Resources Group; a talk on coal procurement at the Sines power plant in Portugal, by Duarte Costa of CPPE-EDP group, Portugal; a presentation by Suresh Iyer of NEPCO on coal at the New England Power Company; and a paper by Giovanni Perego of ENEL Spa, Italy on coal imports and coal use by ENEL.

  8. Unmanned operation of Hydro Power Plants

    International Nuclear Information System (INIS)

    Regula, E.

    2008-01-01

    Intentions to launch unmanned operation are no news, the very first occurred in Hydro Power Plants (HPP) at the time when the first computer technology was implemented into process of power generation, i.e. no later than in 1960 s . ENEL entering Slovenske elektrarne not only revived but significantly accelerated the implementation process of unmanned operation. Experience of ENEL says that unmanned operation means better reliability of the HPP and this is the priority. (author)

  9. US DOE Idaho national laboratory reactor decommissioning

    International Nuclear Information System (INIS)

    Szilagyi, Andrew

    2012-01-01

    , asbestos and mercury among others. Each reactor required isolation in order to be removed. Due to activated metal within the reactor vessels, dose rates above the cores ranged from 50 R/hr to 1200 R/hr. Subsequent dose rates outside the vessels varied from 60 mR/hr to greater than 50 R/hr. Due to the elevated dose rates, the project team decided to fill the ETR and MTR reactor vessels with grout to a level above the core region to reduce dose. To remove the ETR reactor, access to the support shoes was required. These shoes were encased in the high density concrete biological shield approximately 8' below grade. The project team used explosives to remove the biological shield. The demolition had to be controlled to prevent damaging the reactor vessel and to limit the seismic impact on a nearby operating reactor. Upon completion of the blast, the concrete was removed exposing the support shoes for the vessel. Two reactor buildings (ETR and PBF) had to be removed to accommodate lifting systems for the reactor vessels. Two reactors (PBF and MTR) were removed via mobile cranes, two reactors were sized and removed in pieces (ZPPR and MTR), and ETR reactor, due to its weight, was removed via a twin gantry lifting system

  10. Control of reactor coolant flow path during reactor decay heat removal

    International Nuclear Information System (INIS)

    Hunsbedt, A.N.

    1988-01-01

    This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

  11. The bottom-supported fast reactor - system simplifications and enhanced safety

    International Nuclear Information System (INIS)

    Petrozelli, J.; Golan, S.; Kawamura, Yutaka; Kumaoka, Yoshio; Nakagawa, Hiroshi

    1992-01-01

    The 600-MW(electric) bottom-supported fast reactor (BSFR) incorporates the following key features: (1) modular upper internal structure (UIS); (2) electromagnetic pumps (EMPs); (3) low-sodium-void-worth metal-fuel core; and (4) bottom supported reactor vessel (BSRV), which is entirely supported by the basement, except for the control rods, control rod drives (CRDs), UIS, and the stationary plug; by comparison, a top-supported reactor vessel (TSRV) is completely supported by the operating floor. The diameter of the reactor vessel (RV) is 12.8 m (42 ft), and the height (distance from the basemat to the operating floor) is 19.8 m (65 ft). The RV is supported by a single support cylinder anchored to the basemat. The core has 210 driver assemblies and 192 radial blanket assemblies in an annular configuration. The primary heat transport system components consist of four intermediate heat exchangers (IHXs), four EMPs, and four primary reactor auxillary cooling systems. All these components are supported by the BSRV and hang from their tops. Six modular, vertically movable UIS mechanisms clear the UIS from the space over the core during refueling. The top closure is designed to operate at the reactor outlet temperature and is free to expand and contract. Small bellows between the top closure and each UIS model accommodate differential movements and comprise a portion of the cover gas boundary. A 1200-MW(electric) plant with two 600-MW(electric) (twin) nuclear steam supply systems is being studied

  12. Light-water reactor accident classification

    International Nuclear Information System (INIS)

    Washburn, B.W.

    1980-02-01

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

  13. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

    Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

    2000-01-01

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  14. Effect of increase in salinity on ANAMMOX-UASB reactor stability.

    Science.gov (United States)

    Xing, Hui; Wang, Han; Fang, Fang; Li, Kai; Liu, Lianwei; Chen, Youpeng; Guo, Jinsong

    2017-05-01

    The effect of salinity on the anaerobic ammonium oxidation (ANAMMOX) process in a UASB reactor was investigated by analysing ammonium, nitrite, nitrate and TN concentrations, and TN removal efficiency. Extracellular polymeric substances (EPSs) and specific ANAMMOX activity (SAA) were evaluated. Results showed the effluent deteriorated after salinity was increased from 8 to 13 g/L and from 13 to 18 g/L, and TN removal efficiency decreased from 80% to 30% and 80% to 50%, respectively. However, ANAMMOX performance recovered and TN removal efficiency increased to 80% after 40 days when the influent concentrations of [Formula: see text] and [Formula: see text] were 200 mg/L and salinity levels were at 13 and 18 g/L, respectively. The amount of EPSs decreased from 58.9 to 37.1 mg/g volatile suspended solids (VSS) when the reactor was shocked by salinity of 13 g/L, and then increased to 57.2 mg/g VSS when the reactor recovered and ran stably at 13 g/L. The amount of EPSs decreased from 57.2 to 49.1 mg/g VSS when the reactor was shocked by salinity of 18 g/L, and then increased to 60.7 mg/g VSS when the reactor recovered and ran stably at 18 g/L. The amount of EPS and the amounts of polysaccharide, protein and humus showed no evident difference when the reactor recovered from different levels of salinity shocks. Batch tests showed salinity shock load from 8 to 38 g/L inhibited the SAA. However, when the reactor recovered from salinity shocks, SAA was higher compared to that when the reactor was subjected to the same level of salinity shock.

  15. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  16. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  17. Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, J; Park, W S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

  18. Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

  19. MELCOR Severe Accident Analysis on the SMART Reactor

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Jin, Young Ho; Kim, Young In; Kim, Keung Koo; Wang, Ziao; Revankar, Shripad

    2014-01-01

    A severe accident is analyzed for Korea SMR reactor, SMART. Core melt down sequences are analyzed for SMART reactor core using MELCOR version 1.8.5. MELCOR is developed by Sandia National Laboratory for US NRC for the simulation of severe accidents in nuclear power plants. Two cases are simulated here and compared between them; one is the case for core having 3 concentric rings and the other is the case for core having 5 concentric rings. One inch break LOCA scenario is simulated and compared between these two core models. Time sequences for the thermal hydraulic behaviors of RPV and thermal heatup behaviors of reactor core are explained in graphically. Thermal hydraulic behavior such as the change of pressure, level, mass, and temperature of RPV is explained. Thermal heatup behavior of reactor core such as oxidation of cladding, hydrogen generation, core slumping down to lower plenum, and finally creep rupture of PRV lower head is explained. Engineered safety features such as safety injection systems (SIS), and Passive residual heat removal systems (PHRS), etc. are assumed to be not working. One inch break of severe accident is simulated on Korean SMR (SMART) Integral PWR with MELCOR code version 1.8.5. Core melt progression and lower head failure time is very slow compared to other commercial reactors. Simulation on 3 and 5 radial rings core models gives very similar pattern in core cell failure timings. Other various accident scenarios (for example, SBO in Fukushima) will be tried further. Containment behaviors and source term behaviors in severe accident conditions will be analyzed in future

  20. Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

    International Nuclear Information System (INIS)

    Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

    2008-01-01

    .7. Flux distribution measurements, core no. 133 A; 2.8. Increasing of excess reactivity; 2.9; Control rod worth measurement; 2.10. Fuel element reactivity worth in different rings; 2.11. Thermal power calibration; 2.12. Void coefficient measurement; 2.13. Fuel temperature coefficient measurement; 2.14. Fuel temperature measurement in different rings; 3. Pulse Experiments; 4. Conclusions. To summarize, the main purpose of the reactor physics tests was to verify the predicted and calculated values for the most important reactor parameters which had been used in safety evaluation of pulse operation. Comparison of measured results to the calculations show satisfactory agreement so as to qualify the computer codes used and calculational procedures as analytical tools for evaluating the pulse experiments in the future

  1. IAEA activities in the field of research reactors safety

    International Nuclear Information System (INIS)

    Ciuculescu, C.; Boado Magan, H.J.

    2004-01-01

    IAEA activities in the field of research reactor safety are included in the programme of the Division of Nuclear Installations Safety. Following the objectives of the Division, the results of the IAEA missions and the recommendations from International Advisory Groups, the IAEA has conducted in recent years a certain number of activities aiming to enhance the safety of research reactors. The following activities will be presented: (a) the new Requirements for the Safety of Research Reactors, main features and differences with previous standards (SS-35-S1 and SS-35-S2) and the grading approach for implementation; (b) new documents being developed (safety guides, safety reports and TECDOC's); (c) activities related to the Incident Reporting System for Research Reactor (IRSRR); (d) the new features implemented for the INSARR missions; (e) the Code of Conduct on the Safety of Research Reactors adopted by the Board of Governors on 8 March 2004, following the General Conference Resolution GC(45)/RES/10; and (f) the survey on the safety of research reactors published on the IAEA website on February 2003 and the results obtained. (author)

  2. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  3. Computation of nuclear reactor parameters using a stretch Kalman filtering

    International Nuclear Information System (INIS)

    Zwingelstein, G.; Poujol, A.

    1976-01-01

    A method of nonlinear stochastic filtering, the stretched Karman filter, is used for the estimation of two basic parameters involved in the control of nuclear reactor start-up. The corresponding algorithm is stored in a small Multi-8 computer and tested with data recorded for the Ulysse reactor (I.N.S.T.N.). The various practical problems involved in using the algorithm are examined: filtering initialization, influence of the model... The quality and time saving obtained in the computation make it possible for a real time operation, the computer being connected with the reactor [fr

  4. Review of the United Kingdom fast reactor programme - March 1986

    International Nuclear Information System (INIS)

    Bramman, J.I.; John, C.T.; Wheeler, R.C.

    1986-01-01

    The UK programme in the field of fast reactors has continued successfully towards the following main objectives, details of which are contained in subsequent sections of this report: (2) progress with the prototype fast reactor (PFR) which achieved its design power on 4 March 1985; (3) nuclear fuel reprocessing; (4) commercial design studies; (5) structural integrity of LMFBR during its lifetime; (6) R and D work on components of LMFBR; (7) materials study; (8) sodium chemistry; (9) reactor core and fuel design philosophy; (10) safety problems; (11) plant performance studies

  5. Application of the neutron noise technique for measurement of reactivity for subcritical reactor RA-4

    International Nuclear Information System (INIS)

    Orso, J; Marenzana, A

    2012-01-01

    Reactor core RA-4 is divided into two parts that come together to start reactor. The reactor with core separate has the largest subcritical condition, this condition is more secure and therefore the reactor shutdown. In this paper measurements are made of the decay constant of the neutron prompt ' P ', using the α-Rossi and α-Feynman methods to calculate the reactivity of the reactor core for different positions. Both techniques are compared and reactivity is obtained for several position of the reactor core using the α-Rossi technical which is obtained a function that gives the reactivity depending on the separation of the core length. Both techniques are verified using a no multiplicative system. Reactivity values for different position of the core obtained by α-Rossi technique are: $[0 cm] = (-11+/-1) dollar, $[3 cm] = (-7+/-1) dollar, $[3.5 cm] (-5.5+/-0.8) dollar, $[4.2 cm] = (-3.8+/-0.3) dollar y $[4.5] = (-3.0+/-0.1) dollar (author)

  6. Submersion-Subcritical Safe Space (S4) reactor

    International Nuclear Information System (INIS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The Submersion-Subcritical Safe Space (S 4 ) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

  7. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

    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.)

  8. Liquid metal cooled experimental fast reactor simulator

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine; Braz Filho, Francisco; Borges, Eduardo M.; Rosa, Mauricio A.P.; Rocamora, Francisco; Hirdes, Viviane R.

    1997-01-01

    This paper is a continuation of the work that has been done in the area of fast reactor component dynamic analysis, as part of the REARA project at the IEAv/CTA-Brazil. A couple of preceding papers, presented in other meetings, introduced major concept design components of the REARA reactor. The components are set together in order to represent a full model of the power plant. Full model transient results will be presented, together with several parameters to help us to better establish the REARA experimental plant concept. (author). 8 refs., 6 figs., 3 tabs

  9. Desalination of seawater with nuclear reactors

    International Nuclear Information System (INIS)

    Nisan, S.; Volpi, L.

    2003-01-01

    About 40 % of the world population is concerned with water scarcity. This article reviews the different techniques of desalination: distillation (MED and MSF), reverse osmosis (RO), and electrodialysis (ED). The use of nuclear energy rests on several arguments: 1) it is economically efficient compared to fossil energy. 2) nuclear reactors provide heat covering a broad range of temperature, which allows the implementation of all the desalination techniques. 3) the heat normally lost at the heat sink could be used for desalination. And 4) nuclear is respectful of the environment. The feedback experience concerning nuclear desalination is estimated to about 100 reactor-years, it is sufficient to allow the understanding of all the physical and technological processes involved. In Japan, 8 PWR-type reactors are coupled to MED, MSF, and RO desalination techniques, the water produced is used locally mainly for feeding steam generators. (A.C.)

  10. Desalination of seawater with nuclear reactors

    International Nuclear Information System (INIS)

    Nisan, S.; Volpi, L.

    2001-01-01

    About 40 % of the world population is concerned with water scarcity. This article reviews the different techniques of desalination: distillation (MED and MSF), reverse osmosis (RO), and electrodialysis (ED). The use of nuclear energy rests on several arguments: 1) it is economically efficient compared to fossil energy; 2) nuclear reactors provide heat covering a broad range of temperature, which allows the implementation of all the desalination techniques; 3) the heat normally lost at the heat sink could be used for desalination; and 4) nuclear is respectful of the environment. The feedback experience concerning nuclear desalination is estimated to about 100 reactor-years, it is sufficient to allow the understanding of all the physical and technological processes involved. In Japan, 8 PWR-type reactors are coupled to MED, MSF, and RO desalination techniques, the water produced is used locally mainly for feeding steam generators. (A.C.)

  11. High-Uranium-Loaded U3O8-Al fuel element development program. Part 1

    International Nuclear Information System (INIS)

    Martin, M.M.

    1993-01-01

    The High-Uranium-Loaded U 3 O 8 -Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U 3 O 8 -Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U 3 O 8 -Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U 3 O 8 ). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % U involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U 3 O 8 and aluminum

  12. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  13. Proceedings of the 1998 workshop on the utilization of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-10-01

    The 1998 Workshop on the Utilization of Research Reactors, which is the seventh Workshop on the theme of research reactor utilization was held in Yogyakarta and Serpong, Indonesia from February 8 to 14. This Workshop was executed based on the agreement in the Ninth International Conference for Nuclear Cooperation in Asia (ICNCA) held in Tokyo, March 1998. The whole Workshop consists of the Workshop on the theme of following three fields, 1) Neutron Scattering, 2) Neutron Activation analysis and 3) Safe Operation and Maintenance of Research Reactor, and the Sub-workshop carried out the experiment of Neutron Activation analysis. The total number of participants for the workshop was about 100 people from 8 countries, i.e. Australia, China, Indonesia, Korea, Malaysia, Thailand, Vietnam and Japan. The 38 papers are indexed individually. (J.P.N.)

  14. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  15. Nuclear Reactor RA Safety Report, Format and Contents

    International Nuclear Information System (INIS)

    1986-11-01

    This is a new complete version of the safety report of nuclear reactor RA is made according to the recommendations of the IAEA. Report includes all the relevant data needed for evaluation of safe operation of this nuclear facility. Each of seven volumes of this report cover separate topics as follows: (1) introduction; (2) Site characteristics; (3) description of the reactor building and installations; (4) description of the reactor; (5) description of the coolant system; (6) description of the regulation and safety instrumentation; (7) description of the power supply system; (8) description of the auxiliary systems; (9) radiation protection issues; (10) radioactive waste management (11) reactor operation; (12) accident analysis during previous operation; (13) analysis of possible accident causes; (14) safety analysis and preventive actions: (15) analysis of significant accidents; (16) analysis of maximum possible accident; (17) environmental impact analysis in case of accident [sr

  16. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  17. RB Research nuclear reactor RB reactor, Annual report for 2000

    International Nuclear Information System (INIS)

    Milosevic, M.

    2000-12-01

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

  18. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  19. Comparison of advanced reactors program of different international vendors

    International Nuclear Information System (INIS)

    Agnihotri, N.K.

    2001-01-01

    The full text follows. Proposal for presenting a paper on Advanced Reactor Program Given below is the abstract for Track 6 session on Advanced Reactor at the ninth International Conference on Nuclear Engineering being held in Nice, France from April 8. through 12. 2001. This paper will provide an update on Advanced Reactor Program of different vendors in the United States, Japan, and Europe. Specifically the paper will look at the history of different Advanced Reactor Programs, international experience, aspect of economy due to standardization, and the highlights of technical specifications. The paper will also review aspects of Economy due to standardization, public acceptance, required construction time, and the experience of different vendors. The objective of the presentation is to underscore the highlights of the Reactor Program of different vendors in order to keep the attendees of the conference up-to-date. The presentation will be an impartial overview from an outsider's (not part of the Nuclear Steam Supply System's staff). (author)

  20. Fuel technology and performance of non-water cooled reactors. Proceedings of an advisory group meeting held in Vienna, 5-8 December 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-01

    The IAEA Division of Nuclear Fuel Cycle and Waste Management has been closely involved for many years in the collection, analysis and exchange of information relating to the global development of advanced reactor fuel technology and performance. Meetings of experts in this field have been held in 1984 and 1989 and more recently in December 1994 as part of the IAEA`s programme. This publication reviews progress in advanced reactor fuel technology and performance over the past five years, principally related to non-water cooled reactors, namely high temperature gas reactors (HTGRs) and fast reactors (FRs), as well as developments pertaining to thorium fuels and the fuel fabrication technologies. It includes papers from the participants and provides recommendations in key areas where further global co-operation in this field might be usefully initiated or strengthened. The previous two Advisory Group Meetings on Advanced Fuel Technology and Performance, on which separate reports have been published (IAEA-TECDOC-352 (1985) and IAEA-TECDOC-577 (1990)), focused on all types of commercial nuclear reactors. Refs, figs and tabs.

  1. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  2. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  3. Reversed field pinch reactor study 3

    International Nuclear Information System (INIS)

    Hollis, A.A.; Mitchell, J.T.D.

    1977-12-01

    This report, the third of a series on the Reversed Field Pinch Reactor, describes a preliminary concept of the engineering design and layout of this pulsed toroidal reactor, which uses the stable plasma behaviour first observed in ZETA. The basic parameters of the 600 MW(e) reactor are taken from a companion study by Hancox and Spears. The plasma volume is 1.75m minor radius and 16m major radius surrounded by a 1.8m blanket-shield region - with the blanket divided into 14 removable segments for servicing. The magnetic confinement system consists of 28 toroidal field coils situated just outside the blanket and inside the poloidal and vertical field coils and all coils have normal copper conductors. The requirement to incorporate a conducting shell at the front of the blanket to provide a short-time plasma stability has a marked effect on the design. It sets the size of the blanket segment and the scale of the servicing operations, limits the breeding gain and complicates the blanket cooling and its integration with the heat engine. An extensive study will be required to confirm the overall reactor potential of the concept. (author)

  4. Las degradation in a fluidized bed reactor and phylogenetic characterization of the biofilm

    Directory of Open Access Journals (Sweden)

    L. L. Oliveira

    2013-09-01

    Full Text Available A fluidized bed reactor was used to study the degradation of the surfactant linear alkylbenzene sulfonate (LAS. The reactor was inoculated with anaerobic sludge and was fed with a synthetic substrate supplemented with LAS in increasing concentrations (8.2 to 45.8 mg l-1. The removal efficiency of 93% was obtained after 270 days of operation. Subsequently, 16S rRNA gene sequencing and phylogenetic analysis of the sample at the last stage of the reactor operation recovered 105 clones belonging to the domain Bacteria. These clones represented a variety of phyla with significant homology to Bacteroidetes (40%, Proteobacteria (42%, Verrucomicrobia (4%, Acidobacteria (3%, Firmicutes (2%, and Gemmatimonadetes (1%. A small fraction of the clones (8% was not related to any phylum. Such phyla variety indicated the role of microbial consortia in degrading the surfactant LAS.

  5. The activities of the committee 'Kernreaktorregelung' (nuclear reactor control) in the past few years and future projects

    International Nuclear Information System (INIS)

    Knecht, O.

    1976-01-01

    Results achieved so far and future projects are portrayed in detail: 1) VDI/VDE 3527-Graphical symbols for nuclear reactor control; 2) VDI/VDE 3528-Special terms and definitons for nuclear reactor control; 3) 8 data sheets on reactor control; 4) VDI/VDE 3530-Characterisation of reactor control rod drives. (orig./HP) [de

  6. Design requirements, operation and maintenance of gas-cooled reactors

    International Nuclear Information System (INIS)

    1989-06-01

    At the invitation of the Government of the USA the Technical Committee Meeting on Design Requirements, Operation and Maintenance of Gas-Cooled Reactors, was held in San Diego on September 21-23, 1988, in tandem with the GCRA Conference. Both meetings attracted a large contingent of foreign participants. Approximately 100 delegates from 18 different countries participated in the Technical Committee meeting. The meeting was divided into three sessions: Gas-cooled reactor user requirement (8 papers); Gas-cooled reactor improvements to facilitate operation and maintenance (10 papers) and Safety, environmental impacts and waste disposal (5 papers). A separate abstract was prepared for each of these 23 papers. Refs, figs and tabs

  7. Activity report on the utilization of research reactors. Japanese fiscal year, 2003

    International Nuclear Information System (INIS)

    2005-09-01

    During the fiscal year 2003, the Tokai Research Establishment research reactors carried out 8 cycles of joint use reactor operation at JRR-3 and 42 cycles at JRR-4. The research reactors are being utilized for various purposes including experimental studies such as neutron scattering, prompt gamma analysis, neutron radiography and medical irradiation (BNCT), and irradiation utilization such as neutron activation analysis of various samples, Irradiation Test of Reactor Materials and fission track. This volume contains 246 activity reports, which are categorized into the fields of neutron scattering (9 subcategories), neutron radiography, neutron activation analysis, reactor materials, prompt analysis, and others, submitted by the users in JAERI and from other organizations. (author)

  8. Nuclear future: thinking for building. Proceedings of the 12. Brazilian national meeting on reactor physics and thermal hydraulics; 8. General congress on nuclear energy; 5. Brazilian national meeting on nuclear applications

    International Nuclear Information System (INIS)

    2000-01-01

    These proceedings, for the first time, present jointly the 12. Brazilian national meeting on reactor physics and thermal hydraulics (12 ENFIR), 8. General congress on nuclear energy (8. CGEN), and 5. Brazilian national meeting on nuclear applications (5. ENAN). The main theme of discussion was: 'Nuclear Future: thinking for building'. The papers have analysed the progresses of peaceful utilization of nuclear technology and its forecasting for the beginning of the new millennium. The construction of Angra-3 nuclear power plant have been discussed

  9. University Reactor Sharing Program. Final report, September 30, 1992--September 29, 1994

    International Nuclear Information System (INIS)

    Wehring, B.W.

    1995-01-01

    Over the past 20 years, the number of nuclear reactors on university campuses in the US declined from more than 70 to less than 40. Contrary to this trend, The University of Texas at Austin constructed a new reactor facility at a cost of $5.8 million. The new reactor facility houses a new TRIGA Mark II reactor which replaces an in-ground TRIGA Mark I reactor located in a 50-year old building. The new reactor facility was constructed to strengthen the instruction and research opportunities in nuclear science and engineering for both undergraduate and graduate students at The University of Texas. On January 17, 1992, The University of Texas at Austin received a license for operation of the new reactor. Initial criticality was achieved on March 12, 1992, and full power operation, on March 25, 1992. The UT-TRIGA research reactor provides hands-on education, multidisciplinary research and unique service activities for academic, medical, industrial, and government groups. Support by the University Reactor Sharing Programs increases the availability of The University of Texas reactor facility for use by other educational institutions which do not have nuclear reactors

  10. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  11. Nuclear reactor, reactor core thereof, and device for constituting the reactor

    International Nuclear Information System (INIS)

    Takiyama, Masashi.

    1994-01-01

    A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

  12. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

    1979-08-01

    Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

  13. Pellet bed reactor for nuclear thermal propelled vehicles

    International Nuclear Information System (INIS)

    El-Genk, M.; Morley, N.J.; Haloulakos, V.E.

    1991-01-01

    The Pellet Bed Reactor (PeBR) concept is capable of operating at a high power density of up to 3.0 kWt/cu cm and an exit hydrogen gas temperature of 3000 K. The nominal reactor thermal power is 1500 MW and the reactor core is 0.80 m in diameter and 1.3 m high. The nominal PeBR engine generates a thrust of approximately 315 kN at a specific impulse of 1000 s for a mission duration to Mars of 250 days requiring a total firing time of 170 minutes. Because of its low diameter-to-height ratio, PeBR has enough surface area for passive removal of the decay heat from the reactor core. The reactor is equipped with two independent shutdown mechanisms; 8-B4C safety rods and 26 BeO/B4C control drums; each system is capable of operating and scraming the reactor safely. Due to the absence of core internal support structures, the PeBR can be fueled and refueled in orbit using the vacuum of space. These unique features of the PeBR provide for safety during launch, simplicity of handling, deployment, and end-of-life disposal, and vehicle extended lifetime. 11 refs

  14. Fission rate measurements in fuel plate type assembly reactor cores

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1988-01-01

    The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

  15. Power start up of the Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

    1994-01-01

    After accomplishing the physical start-up of the reactor, the power start-up was carried out in February 1984. The power of the reactor has reached: 10 KW on 6/2/1984, 100 KW on 7/2/1984, 200 KW and 300 KW on 8/2/1984; 400 KW and nominal power 500 KW on 9/2/1984. The reactivity temperature coefficient and the xenon poisoning were determined. 3 figs., 12 tabs

  16. Optimal reactor strategy for commercializing fast breeder reactors

    International Nuclear Information System (INIS)

    Yamaji, Kenji; Nagano, Koji

    1988-01-01

    In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

  17. Computerized reactor monitor and control for nuclear reactors

    International Nuclear Information System (INIS)

    Buerger, L.

    1982-01-01

    The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

  18. International Working Group on Fast Reactors Second Annual Meeting. Summary Report

    International Nuclear Information System (INIS)

    1969-01-01

    The Agenda of the Meeting was as follows: Opening of the meeting. 2. Appraisal of the IWGFB's activity for the period from the first annual meeting of the Group. 3. Comments on national programmes on fast breeder reactors. 4. Presentation of general findings and conclusions of national and regional meetings on fast reactor problems held in represented countries and international organisations last year. 5. Comments on the programmes of international meetings on fast reactors to be held in 1969. 6. Consideration of a schedule for meetings on fast reactors in 1970. 7. Suggestions for the topics and location of specialists' meetings in 1969-1970. 8. Suggestions for reviews and studies in the field of fast reactors. 9. The time and place of the third annual meeting of the IWGFR. 10. Closing of the meeting

  19. RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1983

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Raickovic, N.; Radivojevic, J.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1983-12-01

    After regular shutdown in November 1982, inspection of the fuel elements from the RA reactor core which was done from December 1982 - February 1983 has shown that there are deposits of aluminium oxides on the surface of the fuel cladding. After restart The RA reactor was operated at power levels from 1.8 - 2 MW, with 80% enriched uranium dioxide fuel elements. It was found that there was no corrosion of the fuel element cladding and that it was not possible to find the cause of surface deposition on the cladding surfaces without further operation. It was decided to purify the heavy water permanently during operation and to increase the heavy water flow by operating two pumps. This procedure was adopted in order to decrease the possibility of corrosion. The Safety committee of the Institute has approved this procedure for operating the RA reactor in 1983. The core was made of 80% enriched fuel, critical experiments were done until June 1983, and after that the operation was continued at power levels up to 2 MW [sr

  20. Grey water treatment in UASB reactor at ambient temperature.

    Science.gov (United States)

    Elmitwalli, T A; Shalabi, M; Wendland, C; Otterpohl, R

    2007-01-01

    In this paper, the feasibility of grey water treatment in a UASB reactor was investigated. The batch recirculation experiments showed that a maximum total-COD removal of 79% can be obtained in grey-water treatment in the UASB reactor. The continuous operational results of a UASB reactor treating grey water at different hydraulic retention time (HRT) of 20, 12 and 8 hours at ambient temperature (14-24 degrees C) showed that 31-41% of total COD was removed. These results were significantly higher than that achieved by a septic tank (11-14%), the most common system for grey water pre-treatment, at HRT of 2-3 days. The relatively lower removal of total COD in the UASB reactor was mainly due to a higher amount of colloidal COD in the grey water, as compared to that reported in domestic wastewater. The grey water had a limited amount of nitrogen, which was mainly in particulate form (80-90%). The UASB reactor removed 24-36% and 10-24% of total nitrogen and total phosphorus, respectively, in the grey water, due to particulate nutrients removal by physical entrapment and sedimentation. The sludge characteristics of the UASB reactor showed that the system had stable performance and the recommended HRT for the reactor is 12 hours.

  1. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  2. Control of reactor coolant flow path during reactor decay heat removal

    Science.gov (United States)

    Hunsbedt, Anstein N.

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

  3. On blanket concepts of the Helias reactor

    International Nuclear Information System (INIS)

    Wobig, H.; Harmeyer, E.; Herrnegger, F.; Kisslinger, J.

    1999-07-01

    The paper discusses various options for a blanket of the Helias reactor HSR22. The Helias reactor is an upgrade version of the Wendelstein 7-X device. The dimensions of the Helias reactor are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 4.75 T, maximum field 10 T, number of field periods 5, fusion power 3000 MW. The minimum distance between plasma and coils is 1.5 m, leaving sufficient space for a blanket and shield. Three options of a breeding blanket are discussed taking into account the specific properties of the Helias configuration. Due to the large area of the first wall (2600 m 2 ) the average neutron power load on the first wall is below 1 MWm .2 , which has a strong impact on the blanket performance with respect to lifetime and cooling requirements. A comparison with a tokamak reactor shows that the lifetime of first wall components and blanket components in the Helias reactor is expected to be at least two times longer. The blanket concepts being discussed in the following are: the solid breeder concept (HCPB), the dual-coolant Pb-17Li blanket concept and the water-cooled Pb-17Li concept (WCLL). (orig.)

  4. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  5. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  6. Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

    International Nuclear Information System (INIS)

    Hauptman, H.M.; Petro, J.N.; Jacobi, O.

    1995-04-01

    This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

  7. Reactor core for FBR type reactor

    International Nuclear Information System (INIS)

    Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

    1996-01-01

    In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

  8. Space-time reactor kinetics for heterogeneous reactor structure

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1969-11-15

    An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

  9. Post-remedial-action radiological survey of the Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories, Cheswick, Pennsylvania, October 1-8, 1981

    International Nuclear Information System (INIS)

    Flynn, K.F.; Justus, A.L.; Sholeen, C.M.; Smith, W.H.; Wynveen, R.A.

    1984-01-01

    The post-remedial-action radiological assessment conducted by the ANL Radiological Survey Group in October 1981, following decommissioning and decontamination efforts by Westinghouse personnel, indicated that except for the Advanced Fuels Laboratory exhaust ductwork and north wall, the interior surfaces of the Plutonium Laboratory and associated areas within Building 7 and the Advanced Fuels Laboratory within Building 8 were below both the ANSI Draft Standard N13.12 and NRC Guideline criteria for acceptable surface contamination levels. Hence, with the exceptions noted above, the interior surfaces of those areas within Buildings 7 and 8 that were included in the assessment are suitable for unrestricted use. Air samples collected at the involved areas within Buildings 7 and 8 indicated that the radon, thoron, and progeny concentrations within the air were well below the limits prescribed by the US Surgeon General, the Environmental Protection Agency, and the Department of Energy. The Building 7 drain lines are contaminated with uranium, plutonium, and americium. Radiochemical analysis of water and dirt/sludge samples collected from accessible Low-Bay, High-Bay, Shower Room, and Sodium laboratory drains revealed uranium, plutonium, and americium contaminants. The Building 7 drain lines hence are unsuitable for release for unrestricted use in their present condition. Low levels of enriched uranium, plutonium, and americium were detected in an environmental soil coring near Building 8, indicating release or spillage due to Advanced Reactors Division activities or Nuclear Fuel Division activities undr NRC licensure. 60 Co contamination was detected within the Building 7 Shower Room and in soil corings from the environs of Building 7. All other radionuclide concentrations measured in soil corings and the storm sewer outfall sample collected from the environs about Buildings 7 and 8 were within the range of normally expected background concentrations

  10. Nuclear future: thinking for building. Proceedings of the 5. Brazilian national meeting on nuclear applications; 8. General congress on nuclear energy; 12. Brazilian national meeting on reactor physics and thermal hydraulics

    International Nuclear Information System (INIS)

    2000-01-01

    These proceedings, for the first time, present jointly the 12. Brazilian national meeting on reactor physics and thermal hydraulics (12. ENFIR), the 8. General congress on nuclear energy (8. CGEN), and the 5. Brazilian national meeting on nuclear applications (5. ENAN). The main theme of discussion was: 'Nuclear Future: thinking for building'. The papers have analysed the progresses of peaceful utilization of nuclear technology and its forecasting for the beginning of the new millennium. The construction of Angra-3 nuclear power plant have been discussed

  11. Three-dimensional calculations of neutron streaming in the beam tubes of the ORNL HFIR [High Flux Isotope Reactor] Reactor

    International Nuclear Information System (INIS)

    Childs, R.L.; Rhoades, W.A.; Williams, L.R.

    1988-01-01

    The streaming of neutrons through the beam tubes in High Flux Isotope Reactor at Oak Ridge National Laboratory has resulted in a reduction of the fracture toughness of the reactor vessel. As a result, an evaluation of vessel integrity was undertaken in order to determine if the reactor can be operated again. As a part of this evaluation, three-dimensional neutron transport calculations were performed to obtain fluxes at points of interest in the wall of the vessel. By comparing the calculated and measured activation of dosimetry specimens from the vessel surveillance program, it was determined that the calculated flux shape was satisfactory to transpose the surveillance data to the locations in the vessel. A bias factor was applied to correct for the average C/E ratio of 0.69. 8 refs., 7 figs., 3 tabs

  12. Comprehensive study for Anammox process via multistage anaerobic baffled reactors

    Science.gov (United States)

    Ismail, Sherif; Tawfik, Ahmed

    2017-11-01

    Continuous anaerobic ammonia oxidation (Anammox) process in multistage anaerobic baffled (MABR) reactor was investigated. The reactor was operated for approximately 150 days at constant hydraulic retention time (HRT) of 48 h and was fed with synthetic wastewater containing nitrite and ammonium as main substrates. The MABR was inoculated with mixed culture bacteria collected from activated sludge plant (41.6 g MLSS/L and 19.1 g MLVSS/L). The MABR reactor exhibited excellent performance for the start-up of Anammox process within a period of 35 days. The start-up period was divided into four successive phases; cell lysis, lag, activity elevation and steady state. Total inorganic nitrogen (TIN) removal efficiency of 96.8± 0.9% was achieved at steady state conditions, corresponding to nitrogen removal rate (NRR) of 50.2±1.7 mg N/L·d. Moreover, the effect of HRT on the Anammox process was assessed with applying five different HRTs of (48, 38.4, 28.8, 19.2 and 9.6 h). Decreasing HRT from 48 to 9.6 h reduced the removal efficiencies of NH4-N, NO2-N and TIN from 97.7±2.2 to 49.0±9.8%, from 95.7±1.9 to 71.0±8.5% and from 96.8±0.9 to 57.9±9.1%, respectively, that corresponding to reduction in NRR from 50.8±1.2 mg N/L·d at HRT of 48 h to 32.5±5.0 mg N/L·d at HRT of 9.6 h.

  13. An advanced method of heterogeneous reactor theory

    International Nuclear Information System (INIS)

    Kochurov, B.P.

    1994-08-01

    Recent approaches to heterogeneous reactor theory for numerical applications were presented in the course of 8 lectures given in JAERI. The limitations of initial theory known after the First Conference on Peacefull Uses of Atomic Energy held in Geneva in 1955 as Galanine-Feinberg heterogeneous theory:-matrix from of equations, -lack of consistent theory for heterogeneous parameters for reactor cell, -were overcome by a transformation of heterogeneous reactor equations to a difference form and by a development of a consistent theory for the characteristics of a reactor cell based on detailed space-energy calculations. General few group (G-number of groups) heterogeneous reactor equations in dipole approximation are formulated with the extension of two-dimensional problem to three-dimensions by finite Furie expansion of axial dependence of neutron fluxes. A transformation of initial matrix reactor equations to a difference form is presented. The methods for calculation of heterogeneous reactor cell characteristics giving the relation between vector-flux and vector-current on a cell boundary are based on a set of detailed space-energy neutron flux distribution calculations with zero current across cell boundary and G calculations with linearly independent currents across the cell boundary. The equations for reaction rate matrices are formulated. Specific methods were developed for description of neutron migration in axial and radial directions. The methods for resonance level's approach for numerous high-energy resonances. On the basis of these approaches the theory, methods and computer codes were developed for 3D space-time react or problems including simulation of slow processes with fuel burn-up, control rod movements, Xe poisoning and fast transients depending on prompt and delayed neutrons. As a result reactors with several thousands of channels having non-uniform axial structure can be feasibly treated. (author)

  14. Research reactors. Problems of fuel element enrichment reduction. Deliberations and comments

    International Nuclear Information System (INIS)

    1978-10-01

    This paper summarises the main data from the major research reactors in the Federal Republic of Germany utilising highly enriched uranium (HEU) and presently available fuel technology for their fuel elements. The required modification for an adaption of the fabrication to lower enriched fuel are considered as well as the consequences on reactor performance operation and licensing. On the basis of past experience with reactor modifications a rough estimate of 82 months is given for the conversion of a reactor to a modified type of fuel and of 70 months for a fuel test program. The conclusions reflect the own calculations and data from other papers submitted to INFCE-WG 8C

  15. Feasible reactor power cutback logic development for an integral reactor

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  16. The status of fast reactor technology development in China

    International Nuclear Information System (INIS)

    Xu, M.

    2002-01-01

    China, as a developing country with a great number of population and relatively less energy resources, reasonably emphasizes the nuclear energy utilization development. For the long term sustainable energy supply, as for nuclear application the basic strategy of PWR-FBR-Fusion has been settled and envisaged. Due to the economy and experience reasons the nuclear power and technology development with a moderate style are kept in China up to now. In China mainland apart from two NPPs with the total capacity of 2.1 GWe in operation, four NPPs are under construction and two NPPs are planned for the Tenth Five Year Plan (2001-2005). Also another one or two NPPs are still in discussion. It could be foreseen that the total nuclear power capacity will reach 8.5GWe before the year 2005 and 14-15 GWe before 2010 respectively. As the first step for the Chinese fast reactor engineering development the 65MWt China Experimental Fast Reactor (CEFR) is under construction. The main components of primary, secondary and tertiary circuits and of fuel handling system have been ordered. The reactor building under construction has reached 16.8m above the ground. Forty seven components and shielding doors have been installed. It is planned that the construction of reactor building with about 40,000m2 floor surface will be completed in the end of the year 2002 and envisaged that the first criticality of the CEFR will be in the end of 2005. The second step of the Chinese fast reactor engineering development is a 300MWe Prototype Fast Breeder Reactor which is only under consideration up to now. Some important technical selections have been settled, but its design has not yet started. (author)

  17. A hybrid computer simulation of reactor spatial dynamics

    International Nuclear Information System (INIS)

    Hinds, H.W.

    1977-08-01

    The partial differential equations describing the one-speed spatial dynamics of thermal neutron reactors were converted to a set of ordinary differential equations, using finite-difference approximations for the spatial derivatives. The variables were then normalized to a steady-state reference condition in a novel manner, to yield an equation set particularly suitable for implementation on a hybrid computer. One Applied Dynamics AD/FIVE analog-computer console is capable of solving, all in parallel, up to 30 simultaneous differential equations. This corresponds roughly to eight reactor nodes, each with two active delayed-neutron groups. To improve accuracy, an increase in the number of nodes is usually required. Using the Hsu-Howe multiplexing technique, an 8-node, one-dimensional module was switched back and forth between the left and right halves of the reactor, to simulate a 16-node model, also in one dimension. These two versions (8 or 16 nodes) of the model were tested on benchmark problems of the loss-of-coolant type, which were also solved using the digital code FORSIM, with two energy groups and 26 nodes. Good agreement was obtained between the two solution techniques. (author)

  18. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

  19. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  20. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  1. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  2. Calculation and experimental measurements in the Argonauta reactor subcritical and exponential facility

    International Nuclear Information System (INIS)

    Voi, Dante L.; Furieri, Rosane C.A.A.; Renke, Carlos A.C.; Bastos, Wilma S.; Ferreira, Francisco J.O.

    1997-01-01

    Initial measurements were performed on the exponential and subcritical facility installed on the internal thermal column of the Argonauta reactor at IEN-CNEN-Rio de Janeiro, Brazil. The measurements are include in the reactor physics experimental program for integral parameters determination, for both valid and confirmed theoretical models for reactor calculation. Gamma doses and neutron fluxes were measured with telescopic, proportional counters, wire and foil detectors. Experimental data were compared with results obtained by application of CITATION code. (author). 4 refs., 8 figs

  3. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  4. Strategic planning for research reactors. Guidance for reactor managers

    International Nuclear Information System (INIS)

    2001-04-01

    The purpose of this publication is to provide guidance on how to develop a strategic plan for a research reactor. The IAEA is convinced of the need for research reactors to have strategic plans and is issuing a series of publications to help owners and operators in this regard. One of these covers the applications of research reactors. That report brings together all of the current uses of research reactors and enables a reactor owner or operator to evaluate which applications might be possible with a particular facility. An analysis of research reactor capabilities is an early phase in the strategic planning process. The current document provides the rationale for a strategic plan, outlines the methodology of developing such a plan and then gives a model that may be followed. While there are many purposes for research reactor strategic plans, this report emphasizes the use of strategic planning in order to increase utilization. A number of examples are given in order to clearly illustrate this function

  5. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

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

  6. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

  7. Multiregion reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

  8. HYLIFE-II reactor chamber mechanical design

    International Nuclear Information System (INIS)

    House, P.A.

    1992-01-01

    Mechanical design features of the reactor chamber for the HYLIFE-11 inertial confinement fusion power plant are presented. A combination of oscillating and steady, molten salt streams are used for shielding and blast protection. The system is designed for an 8 Hz repetition rate. Beam path clearing, between shots, is accomplished with the oscillating flow. The mechanism for generating the oscillating streams is described. A design configuration of the vessel wall allows adequate cooling and provides extra shielding to reduce thermal stresses to tolerable levels. The bottom portion of the reactor chamber is designed to minimize splash back of the high velocity (20 m/s) salt streams and also recover up to half of the dynamic head

  9. Comparison of reactivity in a flow reactor and a single cylinder engine

    Energy Technology Data Exchange (ETDEWEB)

    Natelson, Robert H.; Johnson, Rodney O.; Kurman, Matthew S.; Cernansky, Nicholas P.; Miller, David L. [Department of Mechanical Engineering and Mechanics, Drexel University, 3141 Chestnut Street, Philadelphia, PA 19104-2875 (United States)

    2010-10-15

    The relative reactivity of 2:1:1 and 1:1:1 mixtures of n-decane:n-butylcyclohexane:n-butylbenzene and an average sample of JP-8 were evaluated in a single cylinder engine and compared to results obtained in a pressurized flow reactor. At compression ratios of 14:1, 15:1, and 16:1, inlet temperature of 500 K, inlet pressure of 0.1 MPa, equivalence ratio of 0.23, and engine speed of 800 RPM, the autoignition delay times were, from shortest to longest, the 2:1:1, followed by the 1:1:1, and then the JP-8. This order corresponded with recent results in a pressurized flow reactor, where the preignition oxidation chemistry was monitored at temperatures of 600-800 K, 0.8 MPa pressure, and an equivalence ratio of 0.30, and where the preignition reactivity from highest to lowest was the 2:1:1, followed by the 1:1:1, and the JP-8. This shows that the relative reactivity at low temperatures in the flow reactor tracks the autoignition tendencies in the engine for these particular fuels. (author) the computed experimental error. (author)

  10. Hydrogen and water reactor safety: proceedings

    International Nuclear Information System (INIS)

    1982-01-01

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

  11. Hydrogen and water reactor safety: proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1982-01-01

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

  12. Considerations in the design of a high power medical isotope production reactor

    International Nuclear Information System (INIS)

    Ball, Russell M.; Nordyke, William H.; Brown, Roy

    2002-01-01

    For the low enriched aqueous homogeneous reactor to be economic in the production of medical isotopes, such as Mo-99 and Sr-89, the power level should be of the order of 100 kWth. This is double the earlier designs and this paper discusses the design changes which must be considered to meet this goal. The topics considered are: 1. Heat removal from the reactor solution; 2. Recombination of radiolytic gases; 3. Adequate radiation shielding; 4. Stability of reactor power with fluctuating reactivity; 5. Adequate cooling of the reflector; 6. Independent shutdown mechanisms; 7. Required volume of the reactor; 8. Economic implementation. (author)

  13. Damage by radiation in structural materials of BWR reactor vessels

    International Nuclear Information System (INIS)

    Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2002-01-01

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

  14. Studies of fragileness in steels of vessels of BWR reactors

    International Nuclear Information System (INIS)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2003-01-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA MARK lll reactor and separately with Ni +3 ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A 2 . (Author)

  15. Känslosvall från vindarnas land / Ann Lingebrandt

    Index Scriptorium Estoniae

    Lingebrandt, Ann

    2008-01-01

    Arvustus: Rooste, Jürgen. Presidenten bor nånstans i närheten / översättning: Peeter Puide. Stockholm : Ellerströms, 2007 ; fs. 2007 / översättning: Peeter Puide. Stockholm : Ellerströms, 2007 ; Kruusa, Kalju. Mötesvis / översättning: Peeter Puide. Stockholm : Ellerströms, 2007 ; Estniska ballader / översättning: Ilmar Laaban. Stockholm : Ellerströms, 2007 ; Laaban, Ilmar. Palingarderomb. Stockholm : Ellerströms, 2007 ; Under, Marie. Gnistor i askan / översättning: Enel Melberg. Stockholm : Ellerströms, 2007 ; Wuolijoki, Hella. Sõngen om kriget / översättning: Enel Melberg. Stockholm : Ellerströms, 2007 ; Alver, Betti. Historien om den vita krõkan / översättning: Enel Melberg. Stockholm : Ellerströms, 2007 ; Kaplinski, Jaan. Nõgonstans vid värlends kant / översättning: Juris Kronbergs och Guntar Godị̧. Tollarp : Ariel skrifter, 2007 ; Kareva, Doris. Tidens gestalt / översättning: Peeter Puide. Stockholm : Ellerströms, 2007

  16. Nuclear reactors situation in Japan after the major earthquake of March 11, 2011. March 12, 2011, 8:00 PM status; Situation des reacteurs nucleaires au Japon suite au seisme majeur servenu le 11 mars 2011. Point de situation du 12 mars 2011 a 20 heures

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    This situation note is established according to the information gained on March 12, 2011, at 8:00 PM, by the crisis centre of the French institute of radiation protection and nuclear safety (IRSN). The situation of the reactors No. 1, 2 and 3 of the Fukushima I site is briefly presented with the progress of the accident management actions. The operation principles of a BWR-type reactor and of a PWR-type reactor are presented in appendix as well as the confinement principle specific to Mark I-type BWR reactors designed by General Electric. The meteorological forecasts of the day are presented in a figure. (J.S.)

  17. Occupational radiation exposure at commercial nuclear power reactors, 1978

    International Nuclear Information System (INIS)

    Brooks, B.G.

    1979-11-01

    An updated compilation is presented of occupational radiation exposures at commercial nuclear power reactors for the years 1969 through 1978. Data received from the 64 light water cooled reactors (LWRs) that had completed at least one year of commercial operation as of December 31, 1978 are included. This represents an increase of seven reactors over the number contained in last year's report. The total number of personnel monitored at LWRs during 1978 increased by approximately 12% to 76,121. The number of workers that received measurable doses, however, increased only 8% to 45,978. The total collective dose for 1978 is estimated to be 31,806 man-rems, a small decrease from last year's value of 32,511, which results in the average dose per worker decreasing slightly to 0.69 rems. The average collective dose per reactor also decreased, by approximately 15%, to a value of 497 man-rems

  18. Anaerobic degradation of coconut husk leachate using UASB-reactor.

    Science.gov (United States)

    Neena, C; Ambily, P S; Jisha, M S

    2007-07-01

    Reffing of coconut husk, the majorprocess in quality coir fibre extraction, causes serious pollution with brackish water lagoons of Kerala. An attempt is made to treat the coconut husk leachate by using a laboratory scale UASB-reactor The experiment was conducted with loading of leachate from 1 kg of fresh coconut husk. The anaerobic treatment was done continuously The parameters like VFA, pH, COD and polyphenols were analysed regularly during the evaluation of the reactor performance. The polyphenol, VFA and COD were diminished gradually with time. The pH of the reactor during the study was found to be in the range of 6-8. The biogas production was increased with loading and about 82% of the total COD/kg husk could be converted to biogas. The maximum polyphenol loading in the reactor was reached to about 298.51 mg/l of husk.

  19. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

    In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  1. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1996-01-01

    The main R and D results of Japanese activities are summarized as follows: (1) the experimental 140 MW(th) sodium cooled fast reactor 'Joyo' provided abundant experimental data and excellent operational records, attaining more than 50,000 hours of operation since its first criticality in 1977; (2) the prototype 280 MW(e) fast reactor 'Monju' reached initial criticality on 5 April 1994; presently Monju is under the cold shutdown state because of secondary sodium leak on 8 December 1995, and multiple cause investigations of the sodium leak are being performed; (3) the Japan Atomic Power Company is promoting design studies for demonstration fast reactor (DFBR) with a power output of 600 MW(e) and R and D for DFBR are being conducted under the cooperation of governmental and private sectors. (author)

  2. Adaptation of GRS calculation codes for Soviet reactors

    International Nuclear Information System (INIS)

    Langenbuch, S.; Petri, A.; Steinborn, J.; Stenbok, I.A.; Suslow, A.I.

    1994-01-01

    The use of ATHLET for incident calculation of WWER has been tested and verified in numerous calculations. Further adaptation may be needed for the WWER 1000 plants. Coupling ATHLET with the 3D nuclear model BIPR-8 for WWER cores clearly improves studies of the influence of neutron kinetics. In the case of FBMK reactors ATHLET calculations show that typical incidents in the complex RMBK reactors can be calculated even though verification still has to be worked on. Results of the 3D-core model QUABOX/CUBBOX-HYCA show good correlation of calculated and measured values in reactor plants. Calculations carried out to date were used to check essential parameters influencing RBMK core behaviour especially dependence of effective voidre activity on the number of control rods. (orig./HP) [de

  3. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    Liu Guisheng

    1995-11-01

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testing of CENDL-2 and ENDF/B-6. (author). 8 refs, 2 figs, 4 tabs

  4. Reactor physics challenges in GEN-IV reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

    2005-02-15

    An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

  5. Reactor physics challenges in GEN-IV reactor design

    International Nuclear Information System (INIS)

    Driscoll, Michael K.; Hejzlar, Pavel

    2005-01-01

    An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

  6. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  7. Reactor physics calculations in the Nordic countries

    International Nuclear Information System (INIS)

    Hoeglund, R.

    1995-01-01

    The seventh biennial meeting on reactor physics calculations in the Nordic countries was arranged by VTT Energy on May 8-9, 1995. 26 papers on different subjects in the field of reactor physics were presented by 45 participants representing research establishments, technical universities, utilities, consultants and suppliers. Resent development and verification of the program systems of ABB Atom, Risoe, Scandpower, Studsvik and VTT Energy were the main topic of the meeting. Benchmarking of the two assembly codes CASMO-4 and HELIOS is proceeding. Cross section data calculated with CASMO-HEX have been validated for the Loviisa reactors. On core analysis ABB atom gives a description on its latest core simulator version POLCA7 with the calculation Core Master 2 and the BWR core supervision system Core Watch. Transient calculations with HEXTRAN, HEXTRAN- PLIM, TRAB, RAMONA, SIMULATE-3K and a code based on PRESTO II/POLCA7 were also presented

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

    International Nuclear Information System (INIS)

    Seifritz, W.

    1983-01-01

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

  9. Licensed operating reactors. Status summary report, data as of September 30, 1984. Volume 8, No. 10

    International Nuclear Information System (INIS)

    1984-11-01

    This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States. The percentage computations, Items 20 through 24 in Section 2, the vendor capacity factors, and actual vs. potential energy production are computed using actual data for the period of consideration. The percentages listed in power generation are computed as an arithmetic average. The factors for the life-span of each unit (the Cumulative column) are reported by the utility. Utility power production data is checked for consistency with previously submitted statistics

  10. A neutron radiography facility on the IRT-2000 reactor

    International Nuclear Information System (INIS)

    Khadduri, I.Y.

    1976-01-01

    A neutron radiography facility has been constructed on the thermal neutron channel of the IRT-2000 reactor. A collimated thermal neutron beam exposure area of 10 cm diameter is obtained with an L/D ratio of 48.8. The film used is cellulose nitrate coated with lithium tetraborate which is insensitive to gamma and beta radiation. Some pictures with good contrast and resolution have been obtained. Pictures of parts of an IRT-2000 reactor fuel pin have also been recorded. (orig) [de

  11. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  12. Laser fusion power reactor system (LFPRS)

    International Nuclear Information System (INIS)

    Kovacik, W.P.

    1977-01-01

    This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements

  13. A digital data acquisition and display system for ITU TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Can, B.; Omuz, S.

    2008-01-01

    Full text: In this study, a digital data acquisition and display system realized for ITU TRIGA Mark-II Reactor is described. This system is realized in order to help the reactor operator and to increase reactor console capacity. The system consists of two main units, which are host computers and RTI-815F, analog devices, data acquisition card. RTI-815F is multi-function analog/digital input/output board that plugs into one of the available long expansion slots in the IBM-PC, PC/XT, PC/AT, or equivalent personal computers. It has 16 analog input channels for single-ended input signals or 8 analog input channels for differential input signals. But its channel capacity can be increased to 32 input channels for single-ended input signals or 16 input channels for differential input signals. RTI-815F board contains 2 analog output channels, 8 digital input channels and 8 digital output channels. In the ITD TRIGA Mark-II Reactor, 6 fuel temperature channels, 3 water temperature channels, 3 control rod position channels and 4 power channels are chosen as analog input signals for RTI-815F. Its digital outputs are assigned to cooling tower fan, primary and secondary pump reactor scram, control rod rundown. During operation, data are automatically archived to disk and displayed on screen. The channel selection time and sampling time can be adjusted. The simulated movement and position of control rods in the reactor core can be noted and displayed. The changes of power, fuel temperature and water temperature can be displayed on the screen as a graphic. In this system both period and reactivity are calculated and displayed on the screen. (authors)

  14. Safe operation and maintenance of research reactor

    International Nuclear Information System (INIS)

    Munsorn, S.

    1999-01-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U 3 O 8 - Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

  15. Safe operation and maintenance of research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Munsorn, S. [Reactor Operation Division, Office of Atomic Energy for Peace, Chatuchak, Bangkok (Thailand)

    1999-10-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U{sub 3}O{sub 8}- Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

  16. Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

    Science.gov (United States)

    Bowles, K. J.

    1974-01-01

    An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

  17. Intrinsically secure fast reactors with dense cores

    International Nuclear Information System (INIS)

    Slessarev, Igor

    2007-01-01

    a small power NPP) then giving an attractive flexibility to the prospective NP. For both above-mentioned ISFR options, which possess the stabilized reactivity at equilibrium, void effects in all reactor types have been favourably corrected: positive void effects in sodium cooled reactors have been radically reduced from 5 to 8$$ (the range of values for the best traditional and some innovative projects) down to $/3. As for the modular core options, all void effects in sodium cooled reactors became negligible. Besides, all void effects in lead cooled ISFR can be 'designed' in a 'harmonious' way: they became modest and favourably negative thus significantly increasing their natural self-protection against severe accidents. These concepts imply using hard neutron spectra in the 'dense' reactor cores identifying fast reactors with: dense fuel (mono-nitride, carbides or metallic alloys similar to known projects like BREST, IFR), elevated fuel in-core fractions (where parasitic neutron capture is significantly depressed), and optimal core-blanket configurations

  18. European supercritical water cooled reactor

    International Nuclear Information System (INIS)

    Schulenberg, T.; Starflinger, J.; Marsault, P.; Bittermann, D.; Maraczy, C.; Laurien, E.; Lycklama a Nijeholt, J.A.; Anglart, H.; Andreani, M.; Ruzickova, M.; Toivonen, A.

    2011-01-01

    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 o C 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

  19. Fuel transfer system for a nuclear reactor

    International Nuclear Information System (INIS)

    Katz, L.R.; Marshall, J.R.; Desmarchais, W.E.

    1977-01-01

    Disclosed is a fuel transfer system for moving nuclear reactor fuel assemblies from a new fuel storage pit to a containment area containing the nuclear reactor, and for transferring spent fuel assemblies under water from the reactor to a spent fuel storage area. The system includes an underwater track which extends through a wall dividing the fuel building from the reactor containment and a car on the track serves as the vehicle for moving fuel assemblies between these two areas. The car is driven by a motor and linkage extending from an operating deck to a chain belt drive on the car. A housing pivotally mounted at its center on the car is hydraulically actuated to vertically receive a fuel assembly which then is rotated to a horizontal position to permit movement through the wall between the containment and fuel building areas. Return to the vertical position provides for fuel assembly removal and the reverse process is repeated when transferring an assembly in the opposite direction. Limit switches used in controlling operation of the system are designed to be replaced from the operating deck when necessary by tools designed for this purpose. 5 claims, 8 figures

  20. Research reactor instrumentation and control technology. Report of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-10-01

    The majority of research reactors operating today were put into operation 20 years ago, and some of them underwent modifications, upgrading and refurbishing since their construction to meet the requirements for higher neutron fluxes. However, a few of these ageing research reactors are still operating with their original instrumentation and control systems (I and C) which are important for reactor safety to guard against abnormal occurrences and reactor control involving startup, shutdown and power regulation. Worn and obsolete I and C systems cause operational problems as well as difficulties in obtaining replacement parts. In addition, satisfying the stringent safety conditions laid out by the nuclear regulatory bodies requires the modernization of research reactors I and C systems and integration of additional instrumentation units to the reactor. In order to clarify these issues and to provide some guidance to reactor operators on state-of-art technology and future trends for the I and C systems for research reactors, a Technical Committee Meeting on Technology and Trends for Research Reactor Instrumentation and Controls was held in Ljubljana, Slovenia, from 4 to 8 December 1995. This publication summarizes the discussions and recommendations resulting from that meeting. This is expected to benefit the research reactor operators planning I and C improvements. Refs, figs, tabs

  1. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

  2. Computerized reactor monitor and control for research reactors

    International Nuclear Information System (INIS)

    Buerger, L.; Vegh, E.

    1981-09-01

    The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

  3. Optimization of design solutions on safety and economy for power unit of NPP with WWER reactor of new generation. Annex 8

    International Nuclear Information System (INIS)

    Krushelnitsky, V.N.; Berkovich, V.M.; Shvyrayev, Yu.; Podshebaykin, A.K.; Fil, N.S.

    2002-01-01

    Development of new generation WWER reactors is being carried out in Russia. These new projects with WWER reactors aim to achieve increased levels of safety and reduced costs. This paper describes these designs and discusses the main factors leading to the safety level increase and the improved economics. (author)

  4. Neutron spectra in two beam ports of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Vega C, H. R.; Hernandez D, V. M.; Aguilar, F.; Paredes, L.; Rivera M, T.

    2013-10-01

    The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a 6 Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

  5. Neutron spectra in two beam ports of the TRIGA Mark III reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R.; Hernandez D, V. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Aguilar, F.; Paredes, L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Rivera M, T., E-mail: fermineutron@yahoo.com [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Unidad Legaria, Av. Legaria 694, 11500 Mexico D. F. (Mexico)

    2013-10-15

    The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a {sup 6}Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

  6. Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

    Energy Technology Data Exchange (ETDEWEB)

    Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

    2016-06-15

    Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

  7. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  8. Reactor. Mind picture of the future Jules-Horowitz Reactor (RHJ)

    International Nuclear Information System (INIS)

    Eustache, S.

    1999-01-01

    This paper gives information about the future research reactor, named Reactor Jules-Horowitz (RJH). This irradiation reactor will be placed at industrialists disposal, for research concerning the competitiveness and the safety french electro-nuclear park. Principles and innovations are detailed. This reactor will respect the ALARA principle (as low as reasonably achievable). (A.L.B.)

  9. Site Investigation for Detection of KIJANG Reactor Core Center

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Tae-Hyun; Kim, Jun Yeon; Kim, Jeeyoung [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    It was planned for the end of March 2017 and extended to April 2018 according to the government budget adjustment. The KJRR project is intended for filling the self-sufficiency of RI demand including Mo-99, increasing the NTD capacity and developing technologies related to the research reactor. In project, site investigation is the first activity that defines seismologic and related geologic aspects of the site. Site investigation was carried out from Oct. 2012 to Jan. 2014 and this study is intended to describe detail procedures in locating the reactor core center. The location of the reactor core center was determined by collectively reviewing not only geological information but also information from architects engineering. EL 50m was selected as ground level by levering construction cost. Four recommended locations (R-1a - R-1d) are displayed for the reactor core center. R-1a was found optimal in consideration of medium rock contour, portion of medium rock covering reactor buildings, construction cost, physical protection and electrical resistivity. It is noted that engineering properties of the medium rock is TCR/RQD 100/53, elastic modulus 7,710 - 8,720MPa, permeability coefficient 2.92E-06cm/s, and S-wave velocity 1,380m/s, sound for foundations of reactor buildings.

  10. SPES-2, an experimental program to support the AP600 development

    Energy Technology Data Exchange (ETDEWEB)

    Tarantini, M. [ENEA, Nuclear Fission Branch, Bolonga (Italy); Medich, C. [SIET S.p.A. Piacenza (Italy)

    1995-09-01

    In support of the development of the AP600 reactor, ENEA, ENEL, ANSALDO and Westinghouse have signed a research agreement. In the framework of this agreement a complex Full Height Full Pressure (FHFP) integral system testing program has been planned on SPES-2 facility. The main purpose of this paper is to point out the status of the test program; describe the hot per-operational test performed and the complete test matrix, giving all the necessary references on the work already published. Two identical Small Break LOCA transients, performed with Pressurizer to Core Make-up Tank (PRZ-CMT) balance line (Test S00203) and without PRZ-CMT balance line (Test S00303) are then compared, to show how the SPES-2 facility can contribute in confirming the new AP600 reactor design choices and can give useful indications to designers. Although the detailed analysis of test data has not been completed, some consideration on the analytical tools utilized and on the SPES-2 capability to simulate the reference plant is then drawn.

  11. Fast reactor development programme in France during 1995

    International Nuclear Information System (INIS)

    Le Rigoleur, C.

    1996-01-01

    In 1995, the total amount of electricity produced in France was 471 TWh, out of which 358.2 TWh (76 %) were produced by nuclear power plants, 36.9 TWh (7.8 %) by conventional thermal plants, and 75.5 TWh (16 %) by hydraulic plants. The net electrical power consumption was 368.7 TWh. At the end of 1995, 'Electricite de France' had 54 PWR units in operation. The availability factor for these units was maintained at 81%. 1995 was marked by a decrease of unexpected shutdowns (1.8% in 1995 instead of 2.2% in 1994), a new reduction in programmed shutdown periods, and a good safety level was maintained. In the field of Fast Reactors, the main events of 1995 were the following. At the end of December 1994, the PHENIX reactor was authorized to perform its 49th cycle at 350 MW th (143 MWe). This 49th cycle was completed without any significant problems on April 7, 1995. During the remainder of the year, the reactor had been shut down in order to carry out several tasks within the scope of the ten-year extension of the PHENIX reactor's lifetime. Concerning the CREYS-MALVILLE plant (SUPER-PHENIX) the first part of the year was devoted to repairing argon leak of one of the IHX. Authorization to restart the reactor was given on August 22. The end of the year was beset by a number of minor incidents. The reactor was restarted at the end of 1995 and reactor power was increased by successive steps (30% Pn (Nominal Power) up to February 6 1996; followed by 50 %...). The 'Decret d'Autorisation de Creation' stipulates that because of its prototype character, SUPER PHENIX will have to be operated under conditions explicitly giving priority to safety and knowledge acquisition, with an objective of research and demonstration. In this context, the so-called 'knowledge acquisition' programme designed to prove the capacity of a large FBR to produce electricity on an industrial scale, to test the consumption of plutonium and minor actinides in a large fast reactor, as well as to provide

  12. Optimising end of generation of Magnox reactors

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  13. The analysis for inventory of experimental reactor high temperature gas reactor type

    International Nuclear Information System (INIS)

    Sri Kuntjoro; Pande Made Udiyani

    2016-01-01

    Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

  14. FBR type reactors

    International Nuclear Information System (INIS)

    Suzuoki, Akira; Yamakawa, Masanori.

    1985-01-01

    Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

  15. FBR type reactor

    International Nuclear Information System (INIS)

    Hayase, Tamotsu.

    1991-01-01

    The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  17. LOFA analyses for the water and helium cooled SEAFP reactors

    International Nuclear Information System (INIS)

    Sponton, L.; Sjoeberg, A.; Nordlinder, S.

    2001-01-01

    This study was performed in the frame of the European long-term fusion safety programme 1999 (SEAFP99). Loss of flow accidents (LOFA) have been studied for two cases, first for a helium cooled reactor with advanced dual-coolant (DUAL) blanket at 100% nominal power. The second case applies to a water-cooled reactor at 20% nominal power. Both transients were simulated with the code MELCOR 1.8.4. The results for the helium cooled reactor show that with a natural circulation flow of helium after the pump stops, the first wall temperature will stay below the temperature for excepted failure of the construction material. For the water cooled reactor, the results show that the pressurizer set point for its liquid volumetric inventory is reached before the plasma facing components attain a critical temperature. The pressurizer set point will induce a plasma shutdown

  18. Training and research reactor facility longevity extension program

    International Nuclear Information System (INIS)

    Carriveau, G.W.

    1991-01-01

    Since 1943, over 550 training and research reactors have been in operation. According to statistics from the International Atomic Energy Agency, ∼325 training and research reactors are currently in service. This total includes a wide variety of designs covering a range of power and research capabilities located virtually around the world. A program has been established at General Atomics (GA) that is dedicated to the support of extended longevity of training and research reactor facilities. Aspects of this program include the following: (1) new instrumentation and control systems; (2) improved and upgraded nuclear monitoring and control channels; (3) facility testing, repair and upgrade services that include (a) pool or tank integrity, (b) cooling system, and (c) water purification system; (4) fuel element testing procedures and replacement; (5) control rod drive rebuilding and upgrades; (6) control and monitoring system calibration and repair service; (7) training services, including reactor operations, maintenance, instrumentation calibration, and repair; and (8) expanded or new uses such as neutron radiography and autoradiography, isotope production, nuclear medicine, activation analysis, and material properties modification

  19. Fast Breeder Reactor Development in France During 1987

    International Nuclear Information System (INIS)

    Asty, M.; Cambillard, E.

    1988-01-01

    On March 8, 1987, a ''sodium leak'' alarm signal was received in the Creys-Malville control room. By the end of March, it had been established beyond all doubt that sodium was, in fact, leaking into the fuel storage drum inter-vessel gap. The reactor has been shut down since May 26. The origin of the leak was located on September 5, after complete drainage of the main tank. Despite the fact that the leak was confined, had had no radiological consequences and cast no doubts on the safe operation of the reactor, the impact of this incident on public opinion, both in France and in the neighbouring European states, was considerable. Two facts would appear to have been decisive. The first was that the reactor had not been shut down immediately, the second was that the leak was only detected and localized in September: it was difficult for people to understand that before its exact position could be determined, certain operations (transfer of a few subassemblies to the reactor core, unloading of the fuel storage drum) had to be performed

  20. Application of non-destructive methods for qualification of the U3O8-Al and U3Si2-Al dispersion fuels in the IEA-R1 Reactor

    International Nuclear Information System (INIS)

    Silva, Jose Eduardo Rosa da

    2011-01-01

    IPEN/CNEN-SP manufactures fuels to be used in its nuclear research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil doesn't have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds, internationally tested and qualified to be used in research reactors, and has gotten experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans to increase the uranium density of these fuels. The objective of this thesis work was to study and to propose a set of non-destructive methods to qualify the dispersions fuels U 3 O 8 -Al e U 3 Si 2 -Al with high uranium density produced at IPEN/CNEN-SP. For that, the irradiation resources in the IEA-R1, and the application of non-destructive methods in the reactor pool available in the Institution were considered. The proposal is to specify, manufacture and irradiate fuel mini plates in IEA-R1 at the maximum densities, qualified internationally, and to monitor their general conditions during the period of irradiation, using non-destructive methods in the reactor pool. In addition to the non-destructive visual inspection and sipping methods, already used at the Institution, the infrastructure for dimensional sub-aquatic testing to evaluate the swelling of irradiated fuel mini plates was completed. The analyses of the results will provide means to assess and decide whether or not to continue with the irradiation of mini plates, until the desired burnup for the irradiation tests at IEA-R1 are reached. (author)

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  2. Licensed operating reactors. Status summary report, data as of 6-30-84. Volume 8, No. 7

    International Nuclear Information System (INIS)

    1984-08-01

    This monthly report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States. The percentage computations, the vendor capacity factors, and actual vs potential energy production are computed using actual data for the period of consideration. The percentages listed in power generation are computed as an arithmetic average. The factors for the life-span of each unit (the Cumulative column) are reported by the utility and are not entirely re-computed by NRC. Utility power production data are checked for consistency with previously submitted statistics

  3. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  4. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  5. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

  6. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  7. Safeguarding on-power fuelled reactors - instrumentation and techniques

    International Nuclear Information System (INIS)

    Waligura, A.; Konnov, Y.; Smith, R.M.; Head, D.A.

    1977-01-01

    Instrumentation and techniques applicable to safeguarding reactors that are fuelled on-power, particularly the CANDU type, have been developed. A demonstration is being carried out at the Douglas Point Nuclear Generating Station in Canada. Irradiated nuclear materials in certain areas - the reactor and spent fuel storage bays - are monitored using photographic and television cameras, and seals. Item accounting is applied by counting spent-fuel bundles during transfer from the reactor to the storage bay and by placing these spent-fuel bundles in a sealed enclosure. Provision is made for inspection and verification of the bundles before sealing. The reactor's power history is recorded by a track-etch power monitor. Redundancy is provided so that the failure of any single piece of equipment does not invalidate the entire safeguards system. Several safeguards instruments and devices have beeen developed and evaluated. These include a super-8 mm surveillance camera system, a television surveillance system, a spent-fuel bundle counter, a device to detect dummy fuel bundles, a cover for enclosing a stack of spent-fuel bundles, and a seal suitable for underwater installation and ultrasonic interrogation. The information provided by these different instruments should increase the effectiveness of Agency safeguards and, when used in combination with other measures, will facilitate inspection at reactor sites

  8. Neutron and gamma characterization within the FFTF reactor cavity

    International Nuclear Information System (INIS)

    Bunch, W.L.; Carter, L.L.; Moore, F.S.; Werner, E.J.; Wilcox, A.D.; Wood, M.R.

    1980-08-01

    Neutron and gamma ray measurements were made within the reactor cavity of the Fast Flux Test Facility (FFTF) to establish the operating characteristics of the Ex-Vessel Flux Monitoring (EVFM) system as a function of reactor power level. A significant effort was made to obtain absolute flux values in order that the measurements could be compared directly with shield design calculations. Good agreement was achieved for neutrons and for both the prompt and delayed components of the gamma ray field. 8 figures, 3 tables

  9. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

  10. Nuclear reactor instrumentation at research reactor renewal

    International Nuclear Information System (INIS)

    Baers, B.; Pellionisz, P.

    1981-10-01

    The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

  11. WRAP: a water reactor analysis package

    International Nuclear Information System (INIS)

    Anderson, M.M.

    1977-06-01

    The modular computational system known as the Water Reactor Analysis Package (WRAP) has been developed at the Savannah River Laboratory. WRAP is essentially a reprogrammed version of the RELAP4 computer code with an extensively restructured input format, a dynamic dimensioning capability and additional computational capabilities such as an automatic steady-state option for pressurized water reactors and an automatic restart capability with provision for renodalization. The report describes the capabilities of WRAP at its current stage of development. The addition of new capabilities (e.g., a BWR steady-state capability), the inclusion of improved models (e.g., models in RELAP4/M0D8) and the development of improved numerical techniques to reduce execution time are being planned at this time

  12. Assessment of core protection and monitoring systems for an advanced reactor SMART

    International Nuclear Information System (INIS)

    In, Wang Kee; Hwang, Dae Hyun; Yoo, Yeon Jong; Zee, Sung Qunn

    2002-01-01

    Analogue and digital core protection/monitoring systems were assessed for the implementation in an advanced reactor. The core thermal margins to nuclear fuel design limits (departure from nucleate boiling and fuel centerline melting) were estimated using the design data for a commercial pressurized water reactor and an advanced reactor. The digital protection system resulted in a greater power margin to the fuel centerline melting by at least 30% of rated power for both commercial and advanced reactors. The DNB margin with the digital system is also higher than that for the analogue system by 8 and 12.1% of rated power for commercial and advanced reactors, respectively. The margin gain with the digital system is largely due to the on-line calculations of DNB ratio and peak local power density from the live sensor signals. The digital core protection and monitoring systems are, therefore, believed to be more appropriate for the advanced reactor

  13. NOVEL REACTOR FOR THE PRODUCTION OF SYNTHESIS GAS

    Energy Technology Data Exchange (ETDEWEB)

    Vasilis Papavassiliou; Leo Bonnell; Dion Vlachos

    2004-12-01

    Praxair investigated an advanced technology for producing synthesis gas from natural gas and oxygen This production process combined the use of a short-reaction time catalyst with Praxair's gas mixing technology to provide a novel reactor system. The program achieved all of the milestones contained in the development plan for Phase I. We were able to develop a reactor configuration that was able to operate at high pressures (up to 19atm). This new reactor technology was used as the basis for a new process for the conversion of natural gas to liquid products (Gas to Liquids or GTL). Economic analysis indicated that the new process could provide a 8-10% cost advantage over conventional technology. The economic prediction although favorable was not encouraging enough for a high risk program like this. Praxair decided to terminate development.

  14. Reactor System Design

    International Nuclear Information System (INIS)

    Chi, S. K.; Kim, G. K.; Yeo, J. W.

    2006-08-01

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  15. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1983-01-01

    Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

  16. Prevention device for rapid reactor core shutdown in BWR type reactors

    International Nuclear Information System (INIS)

    Koshi, Yuji; Karatsu, Hiroyuki.

    1986-01-01

    Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

  17. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  18. The International Thermonuclear Experimental Reactor configuration evolution

    International Nuclear Information System (INIS)

    Lousteau, D.C.; Nelson, B.E.; Lee, V.D.; Thomson, S.L.; Miller, J.M.; Lindquist, W.B.

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) conceptual design activities consist of two phases: a definition phase, completed in September 1988, and a design phase, now in progress. The definition phase was successful in identifying a consistent set of technical characteristics and the broad definition of the required reactor configuration and hardware. Scheduled for completion in November 1990, the design phase is producing a more detailed definition of the required components, a first cost estimate, and a description of site requirements. A major activity in the ITER design phase is the period of joint work conducted at the Max Planck Institute for Plasma Physics, Garching, Federal Republic of Germany, from June through October 1989. An official report of the findings and conclusions of this activity will be submitted to and published by the International Atomic Energy Agency (IAEA). This paper highlights the evolution of the reactor mechanical configuration since the conclusion of the definition phase. 8 figs., 2 tabs

  19. Accident Source Terms for Pressurized Water Reactors with High-Burnup Cores Calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Goldmann, Andrew; Kalinich, Donald A.; Powers, Dana A.

    2016-12-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs 2 MoO 4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  20. High-temperature and breeder reactors - economic nuclear reactors of the future

    International Nuclear Information System (INIS)

    Djalilzadeh, A.M.

    1977-01-01

    The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

  1. Heterogeneous reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

  2. Proceedings of first SWCR-KURRI academic seminar on research reactors and related research topics

    International Nuclear Information System (INIS)

    Kimura, Itsuro; Cong, Zhebao

    1986-01-01

    These are the proceedings of an academic seminar on research reactors and related research topics held at the Southwest Centre for Reactor Engineering Research and Design in Chengdu, Sichuan, People's Republic of China in September 24-26 in 1985. Included are the chairmen's addresses and 10 papers presented at the seminar in English. The titles of these papers are: (1) Nuclear Safety and Safeguards, (2) General Review of Thorium Research in Japanese Universities, (3) Comprehensive Utilization and Economic Analysis of the High Flux Engineering Test Reactor, (4) Present States of Applied Health Physics in Japan, (5) Neutron Radiography with Kyoto University Reactor, (6) Topics of Experimental Works with Kyoto University Reactor, (7) Integral Check of Nuclear Data for Reactor Structural Materials, (8) The Reactor Core, Physical Experiments and the Operation Safety Regulation of the Zero Energy Thermal Reactor for PWR Nuclear Power Plant, (9) HFETR Core Physical Parameters at Power, (10) Physical Consideration for Loads of Operated Ten Cycles in HFETR. (author)

  3. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  4. Molten salt reactor technology for long-range and wide-scale nuclear energy system

    International Nuclear Information System (INIS)

    Ignatiev, V.; Alexseev, P.; Menshikov, L.; Prusakov, V.; Subbotine, S.

    1997-01-01

    A possibility of creation of multi-component nuclear power system in which alongside with thermal and fast reactors, molten salt burner reactors, for incineration of weapon grade plutonium, some minor actinides and transmutation of some fission products will be presented. The purposes of this work are to review the present status of the molten salt reactor technology and innovative non-aqueous chemical processing methods, to indicate the importance of the uncertainties remaining, to identify the additional work needed, and to evaluate the probability of success in obtaining improved safety characteristics for new concept of molten salt - burner reactor with external neutron source. 8 refs., 3 figs., 2 tabs

  5. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

  6. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

    Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

  7. EBR-II [Experimental Breeder Reactor-II] system surveillance using pattern recognition software

    International Nuclear Information System (INIS)

    Mott, J.E.; Radtke, W.H.; King, R.W.

    1986-02-01

    The problem of most accurately determining the Experimental Breeder Reactor-II (EBR-II) reactor outlet temperature from currently available plant signals is investigated. Historically, the reactor outlet pipe was originally instrumented with 8 temperature sensors but, during 22 years of operation, all these instruments have failed except for one remaining thermocouple, and its output had recently become suspect. Using pattern recognition methods to compare values of 129 plant signals for similarities over a 7 month period spanning reconfiguration of the core and recalibration of many plant signals, it was determined that the remaining reactor outlet pipe thermocouple is still useful as an indicator of true mixed mean reactor outlet temperature. Application of this methodology to investigate one specific signal has automatically validated the vast majority of the 129 signals used for pattern recognition and also highlighted a few inconsistent signals for further investigation

  8. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    Powell, J.R.

    1977-01-01

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

  9. Fusion-fission hybrid reactors

    International Nuclear Information System (INIS)

    Greenspan, E.

    1984-01-01

    This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

  10. Proceedings of the 1992 topical meeting on advances in reactor physics

    International Nuclear Information System (INIS)

    1992-01-01

    This document, Volume 1, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Code Benchmarks and Validation; Fuel Management; Nodal Methods for Diffusion Theory; Criticality Safety and Applications and Waste; Core Computational Systems; Nuclear Data; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual papers have been cataloged separately. (FI)

  11. Reactor protection systems for the Replacement Research Reactor, ANSTO

    International Nuclear Information System (INIS)

    Morris, C.R.

    2003-01-01

    The 20-MW Replacement Research Reactor Project which is currently under construction at ANSTO will have a combination of a state of the art triplicated computer based reactor protection system, and a fully independent, and diverse, triplicated analogue reactor protection system, that has been in use in the nuclear industry, for many decades. The First Reactor Protection System (FRPS) consists of a Triconex triplicated modular redundant system that has recently been approved by the USNRC for use in the USA?s power reactor program. The Second Reactor Protection System is a hardwired analogue system supplied by Foxboro, the Spec 200 system, which is also Class1E qualified. The FRPS is used to drop the control rods when its safety parameter setpoints have been reached. The SRPS is used to drain the reflector tank and since this operation would result in a reactor poison out due to the time it would take to refill the tank the FRPS trip setpoints are more limiting. The FRPS and SRPS have limited hardwired indications on the control panels in the main control room (MCR) and emergency control centre (ECC), however all FRPS and SRPS parameters are capable of being displayed on the reactor control and monitoring system (RCMS) video display units. The RCMS is a Foxboro Series I/A control system which is used for plant control and monitoring and as a protection system for the cold neutron source. This paper will provide technical information on both systems, their trip logics, their interconnections with each other, and their integration into the reactor control and monitoring system and control panels. (author)

  12. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

    A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

  13. European blanket development for a demo reactor

    International Nuclear Information System (INIS)

    Giancarli, L.; Proust, E.; Anzidei, L.

    1994-01-01

    There are four breeding blanket concepts for a fusion DEMO reactor under development within the framework of the fusion technology programme of the European Union (EU). This paper describes the design of these concepts, the accompanying R + D programme and the status of the development. (authors). 8 figs., 1 tab

  14. A multi-purpose reactor

    International Nuclear Information System (INIS)

    Changwen Ma

    2000-01-01

    An integrated natural circulation self pressurized reactor can be used for sea water desalination, electrogeneration, ship propulsion and district or process heating. The reactor can be used for ship propulsion because it has following advantages: it is a integrated reactor. Whole primary loop is included in a size limited pressure vessel. For a 200 MW reactor the diameter of the pressure vessel is about 5 m. It is convenient to arranged on a ship. Hydraulic driving facility of control rods is used on the reactor. It notably decreases the height of the reactor. For ship propulsion, smaller diameter and smaller height are important. Besides these, the operation reliability of the reactor is high enough, because there is no rotational machine (for example, circulating pump) in safety systems. Reactor systems are simple. There are no emergency water injection system and boron concentration regulating system. These features for ship propulsion reactor are valuable. Design of the reactor is based on existing demonstration district heating reactor design. The mechanic design principles are the same. But boiling is introduced in the reactor core. Several variants to use the reactor as a movable seawater desalination plant are presented in the paper. When the sea water desalination plant is working to produce fresh water, the reactor can supply electricity at the same time to the local electricity network. Some analyses for comprehensive application of the reactor have been done. Main features and parameters of the small (Thermopower 200 MW) reactor are given in the paper. (author)

  15. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  16. Dismantling of the reactor block of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Cremer, J. [Siempelkamp Nukleartechnik GmbH, Heidelberg (Germany)

    2003-07-01

    By the end of 1998 the complete secondary cooling system and the major part of the primary cooling system were dismantled. Furthermore, the experimental devices, including a rabbit system conceived as an in-core irradiation device, were disassembled and disposed of. In total, approx. 65 t of contaminated and/or activated material as well as approx. 70 t of clearance-measured material were disposed of within the framework of these activities. The dismantling of the coolant loops and experimental devices was followed in 2000 by the removal of the reactor tank internals and the subsequent draining of the reactor tank water. The reactor tank internals were essentially the core support plate, the core box, the flow channel and the neutron flux bridges (s. Fig. 2, detailed reactor core). All components consisted of aluminium, the connecting elements such as bolts and nuts, however, of stainless steel. Due to the high activation of the core internals, disassembly had to be remotely controlled under water. All removal work was carried out from a tank intermediate floor (s. Fig. 2). These activities, which served for preparing the dismantling of the reactor block, were completed in summer 2001. The waste parts arising were transferred to the Service Department for Decontamination of the Research Centre. This included approx. 2.5 t of waste parts with a total activity of approx. 8 x 10{sup 11} Bq. (orig.)

  17. Supercritical Water Reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

    2014-01-01

    The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

  18. Tank type reactor

    International Nuclear Information System (INIS)

    Otsuka, Fumio.

    1989-01-01

    The present invention concerns a tank type reactor capable of securing reactor core integrity by preventing incorporation of gases to an intermediate heat exchanger, thgereby improving the reliability. In a conventional tank type reactor, since vortex flows are easily caused near the inlet of an intermediate heat exchanger, there is a fear that cover gases are involved into the coolant main streams to induce fetal accidents. In the present invention, a reactor core is suspended by way of a suspending body to the inside of a reactor vessel and an intermediate heat exchanger and a pump are disposed between the suspending body and the reactor vessel, in which a vortex current preventive plate is attached at the outside near the coolant inlet on the primary circuit of the intermediate heat exchanger. In this way vortex or turbulence near the inlet of the intermediate heata exchanger or near the surface of coolants can be prevented. Accordingly, the cover gases are no more involved, to insure the reactor core integrity and obtain a tank type nuclear reactor of high reliability. (I.S.)

  19. New about research reactors

    International Nuclear Information System (INIS)

    Egorenkov, P.M.

    2001-01-01

    The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru

  20. Westinghouse Small Modular Reactor nuclear steam supply system design

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

  1. Biological oxidation of dissolved methane in effluents from anaerobic reactors using a down-flow hanging sponge reactor.

    Science.gov (United States)

    Hatamoto, Masashi; Yamamoto, Hiroki; Kindaichi, Tomonori; Ozaki, Noriatsu; Ohashi, Akiyoshi

    2010-03-01

    Anaerobic wastewater treatment plants discharge dissolved methane, which is usually not recovered. To prevent emission of methane, which is a greenhouse gas, we utilized an encapsulated down-flow hanging sponge reactor as a post-treatment to biologically oxidize dissolved methane. Within 3 weeks after reactor start-up, methane removal efficiency of up to 95% was achieved with a methane removal rate of 0.8 kg COD m(-3) day(-1) at an HRT of 2 h. After increasing the methane-loading rate, the maximum methane removal rate reached 2.2 kg COD m(-3) day(-1) at an HRT of 0.5 h. On the other hand, only about 10% of influent ammonium was oxidized to nitrate during the first period, but as airflow was increased to 2.5 L day(-1), nitrification efficiency increased to approximately 70%. However, the ammonia oxidation rate then decreased with an increase in the methane-loading rate. These results indicate that methane oxidation occurred preferentially over ammonium oxidation in the reactor. Cloning of the 16S rRNA and pmoA genes as well as phylogenetic and T-RFLP analyses revealed that type I methanotrophs were the dominant methane oxidizers, whereas type II methanotrophs were detected only in minor portion of the reactor. Copyright 2009 Elsevier Ltd. All rights reserved.

  2. Project RA Research nuclear reactor - Annual report 1993 with comparative review for the period 1991 - 1993

    International Nuclear Information System (INIS)

    1993-12-01

    Research reactor RA Annual report for year 1993 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. The ninth separate annex deals with the feasibility of RA reactor applications. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data [sr

  3. Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

    International Nuclear Information System (INIS)

    Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

    2013-01-01

    The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

  4. Design guide for Category III reactors: pool type reactors

    International Nuclear Information System (INIS)

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

    1978-11-01

    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, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems

  5. The computerized reactor period measurement system for China fast burst reactor-II

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1996-01-01

    The article simply introduces the hardware, principle, and software of the computerized reactor period measurement system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between fission yield and pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computerized measurement system makes the reactor period measurement into automatical and intelligent and also improves the speed and precision of period data on-line process

  6. Computer measurement system of reactor period for China fast burst reactor-II

    International Nuclear Information System (INIS)

    Zhao Wuwen; Jiang Zhiguo

    1997-01-01

    The author simply introduces the hardware, principle, and software of the reactor period computer measure system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between Fission yield and Pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computer measure system makes the reactor period measurement into automation and intellectualization and also improves the speed and precision of period data process on-line

  7. Reactor safety method

    International Nuclear Information System (INIS)

    Vachon, L.J.

    1980-01-01

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

  8. Fast Pyrolysis of Lignin Using a Pyrolysis Centrifuge Reactor

    DEFF Research Database (Denmark)

    Trinh, Ngoc Trung; Jensen, Peter Arendt; Sárossy, Zsuzsa

    2013-01-01

    Fast pyrolysis of lignin from an ethanol plant was investigated on a lab scale pyrolysis centrifuge reactor (PCR) with respect to pyrolysis temperature, reactor gas residence time, and feed rate. A maximal organic oil yield of 34 wt % dry basis (db) (bio-oil yield of 43 wt % db) is obtained...... at temperatures of 500−550 °C, reactor gas residence time of 0.8 s, and feed rate of 5.6 g/min. Gas chromatography mass spectrometry and size-exclusion chromatography were used to characterize the Chemical properties of the lignin oils. Acetic acid, levoglucosan, guaiacol, syringols, and p-vinylguaiacol are found...... components and molecular mass distribution of the lignin oils. The obtained lignin oil has a very different components composition when compared to a beech wood oil....

  9. Small propulsion reactor design based on particle bed reactor concept

    International Nuclear Information System (INIS)

    Ludewig, H.; Lazareth, O.; Mughabghab, S.; Perkins, K.; Powell, J.R.

    1989-01-01

    In this paper Particle Bed Reactor (PBR) designs are discussed which use 233 U and /sup 242m/Am as fissile materials. A constant total power of 100MW is assumed for all reactors in this study. Three broad aspects of these reactors is discussed. First, possible reactor designs are developed, second physics calculations are outlined and discussed and third mass estimates of the various candidates reactors are made. It is concluded that reactors with a specific mass of 1 kg/MW can be envisioned of 233 U is used and approximately a quarter of this value can be achieved if /sup 242m/Am is used. If this power level is increased by increasing the power density lower specific mass values are achievable. The limit will be determined by uncertainties in the thermal-hydraulic analysis. 5 refs., 5 figs., 6 tabs

  10. Code of Conduct on the Safety of Research Reactors

    International Nuclear Information System (INIS)

    2006-09-01

    The Board of Governors of the International Atomic Energy Agency (IAEA) adopted the Code of Conduct on the Safety of Research Reactors on 8 March 2004. The Board's action was the culmination of several years of work to develop the Code and obtain a consensus on its provisions. The process leading to the Code began in 1998, when the International Nuclear Safety Advisory Group (INSAG) informed the Director General of concerns about the safety of research reactors. In 2000, INSAG recommended that the Secretariat begin developing an international protocol or a similar legal instrument to address those concerns. In September 2000, in resolution GC(44)/RES/14, the General Conference requested the Secretariat ''within its available resources, to continue work on exploring options to strengthen the international nuclear safety arrangements for civil research reactors, taking due account of input from INSAG and the views of other relevant bodies''. A working group convened by the Secretariat pursuant to that request recommended that ''the Agency consider establishing an international action plan for research reactors'' and that the action plan include preparation of a Code of Conduct ''that would clearly establish the desirable attributes for management of research reactor safety''. In September 2001, the Board requested that the Secretariat develop and implement, in conjunction with Member States, an international research reactor safety enhancement plan which included preparation of a Code of Conduct on the Safety of Research Reactors. Subsequently, in resolution GC(45)/RES/10.A, the General Conference endorsed the Board's request. Pursuant to that request, a Code of Conduct on the Safety of Research Reactors was drafted at two meetings of an Open-ended Working Group of Legal and Technical Experts. This draft Code of Conduct was circulated to all Member States for comment. On the basis of the responses received, a revised draft of the Code was prepared by the Secretariat

  11. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    Baeten, Peter

    2006-01-01

    This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

  12. Cobalt-60 control in Ontario Hydro reactors

    International Nuclear Information System (INIS)

    Lacy, C.S.

    1988-01-01

    This paper discusses the impact of specifying reduced Cobalt-59 in the primary heat transport circuit materials of construction on the radiation fields developed around the primary circuit. An eight-fold reduction in steam generator radiation fields due to Cobalt-60 has been observed for two identical sets of reactors, one with and one without Cobalt-59 control. The comparison is between eight reactors at the Pickering Nuclear Generating Station (PNGS). Units 5 to 8 (PNGS-B) are identical to Units 1 to 4 (PNGS-A) except that PNGS-B has reduced impurity Cobalt-59 in the alloys of construction and a reduced use of stellite. The effects of chemistry control are also discussed

  13. Safety of light water reactors. Risks of nuclear technology

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  14. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  15. High-Sensitivity Fast Neutron Detector KNK-2-8M

    Science.gov (United States)

    Koshelev, A. S.; Dovbysh, L. Ye.; Ovchinnikov, M. A.; Pikulina, G. N.; Drozdov, Yu. M.; Chuklyaev, S. V.; Pepyolyshev, Yu. N.

    2017-12-01

    The design of the fast neutron detector KNK-2-8M is outlined. The results of he detector study in the pulse counting mode with pulses from 238U nuclei fission in the radiator of the neutron-sensitive section and in the current mode with separation of functional section currents are presented. The possibilities of determination of the effective number of 238U nuclei in the radiator of the neutron-sensitive section are considered. The diagnostic capabilities of the detector in the counting mode are demonstrated, as exemplified by the analysis of reference data on characteristics of neutron fields in the BR-1 reactor hall. The diagnostic capabilities of the detector in the current mode are demonstrated, as exemplified by the results of measurements of 238U fission intensity in the power startup of the BR-K1 reactor in the fission pulse generation mode with delayed neutrons and the detector placed in the reactor cavity in conditions of large-scale variation of the reactor radiation fields.

  16. Guide to power reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-15

    The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

  17. Analysis of dynamic stability and safety of reactor system by reactor simulator

    International Nuclear Information System (INIS)

    Raisic, N.

    1963-11-01

    In order to enable qualitative analysis of dynamic properties of reactors RA and RB, mathematical models of these reactors were formulated and adapted for solution on analog computer. This report contains basic assessments for creating the model and complete equations for each reactor. Model was used to analyse three possible accidents at the RA reactor and possible hypothetical accidents at the RB reactor

  18. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  19. Medium temperature carbon dioxide gas turbine reactor

    International Nuclear Information System (INIS)

    Kato, Yasuyoshi; Nitawaki, Takeshi; Muto, Yasushi

    2004-01-01

    A carbon dioxide (CO 2 ) gas turbine reactor with a partial pre-cooling cycle attains comparable cycle efficiencies of 45.8% at medium temperature of 650 deg. C and pressure of 7 MPa with a typical helium (He) gas turbine reactor of GT-MHR (47.7%) at high temperature of 850 deg. C. This higher efficiency is ascribed to: reduced compression work around the critical point of CO 2 ; and consideration of variation in CO 2 specific heat at constant pressure, C p , with pressure and temperature into cycle configuration. Lowering temperature to 650 deg. C provides flexibility in choosing materials and eases maintenance through the lower diffusion leak rate of fission products from coated particle fuel by about two orders of magnitude. At medium temperature of 650 deg. C, less expensive corrosion resistant materials such as type 316 stainless steel are applicable and their performance in CO 2 have been proven during extensive operation in AGRs. In the previous study, the CO 2 cycle gas turbomachinery weight was estimated to be about one-fifth compared with He cycles. The proposed medium temperature CO 2 gas turbine reactor is expected to be an alternative solution to current high-temperature He gas turbine reactors

  20. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  1. RA reactor operation and maintenance, Annual report 1974

    International Nuclear Information System (INIS)

    Milosevic, D. et al

    1974-12-01

    During 1974, RA reactor was operated at nominal power for 194 days and 13 days at lower power levels. The total production was 30711 MWh which is 2.4% higher than planned. Practically there were no discrepancies from the plan. The reactor was used for irradiation and experiments according to the demand of 437 users. This report contains detailed data about reactor power and experiments performed in 1974. Total number of afety shutdowns was 11, of which 8 were caused by power cuts, and 3 due to human error. Maximum individual personnel exposure dose was 50% of the maximum permissible dose. There were no accidents during this year. Decontamination of surfaces was less than during previous years. About 805 m 2 of surfaces and 178 objects were decontaminated. It was concluded that the successful operation in 1974 has a special significance taking into account the financial problems [sr

  2. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm 3 was by then in routine use, illustrated how far work has progressed

  3. Sodium-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hammers, H.W.

    1982-01-01

    The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

  4. Irradiation experience of IPEN fuel at IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

    1998-01-01

    IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

  5. Atomic reactor thermal engineering

    International Nuclear Information System (INIS)

    Kim, Gwang Ryong

    1983-02-01

    This book starts the introduction of atomic reactor thermal engineering including atomic reaction, chemical reaction, nuclear reaction neutron energy and soon. It explains heat transfer, heat production in the atomic reactor, heat transfer of fuel element in atomic reactor, heat transfer and flow of cooler, thermal design of atomic reactor, design of thermodynamics of atomic reactor and various. This deals with the basic knowledge of thermal engineering for atomic reactor.

  6. Experiment on continuous operation of the Brazilian IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Freitas Pintaud, M. de

    1994-01-01

    In order to increase the radioisotope production in the IEA-R1 research reactor at IPEN/CNEN-SP, it has been proposed a change in its operation regime from 8 hours per day and 5 days per week to continuous 48 hours per week. The necessary reactor parameters for this new operation regime were obtained through an experiment in which the reactor was for the first time operated in the new regime. This work presents the principal results from this experiment: xenon reactivity, new shutdown margins, and reactivity loss due to fuel burnup in the new operation regime. (author)

  7. Measured and calculated effective delayed neutron fraction of the IPR-R1 Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rose Mary G.P.; Dalle, Hugo M.; Campolina, Daniel A.M., E-mail: souzarm@cdtn.b, E-mail: dallehm@cdtn.b, E-mail: campolina@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The effective delayed neutron fraction, {beta}{sub eff}, one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U{sup 235}. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the {beta}{sub eff} value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current {beta}{sub eff} parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 {+-} 38) pcm and (744 {+-} 20) pcm, respectively. The {beta}{sub eff} was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. (author)

  8. Damage analysis of TRIGA MARK II Bandung reactor tank material structure

    International Nuclear Information System (INIS)

    Soedardjo; Sumijanto

    2000-01-01

    Damage of Triga Mark II Bandung reactor tank material structure has been analyzed. The analysis carried out was based on ultrasonic inspection result in 1996 and the monthly reports of reactor operation by random data during 1988 up to 1995. Ultrasonic test data had shown that thinning processes on south and west region of reactor out side wall at upper part of water level had happened. Reactor operation data had shown the demineralized water should be added monthly to the reactor and bulk shielding water tank. Both reactor and bulk shielding tank are shielded by concrete of Portland type I cement consisting of CaO content about 58-68 %. The analysis result shows that the reaction between CaO and seepage water from bulk shielding wall had taken place and consequently the reactor out sidewall surroundings became alkaline. Based on Pourbaix diagram, the aluminum reactor tank made of aluminum alloy 6061 T6 would be corroded easily at pH equal an greater than 8.6. The passive layer AI 2 O 3 aluminum metal surface would be broken due to water reaction taken place continuously at high pH and produces hydrogen gas. The light hydrogen gas would expand the concrete cement and its expanding power would open the passive layer of aluminum metal upper tank. The water sea pages from adding water into reactor tank could indicate the upper water level tank corrosion is worse than the lower water level tank. (author)

  9. Current status of the Thai Research Reactor (TRR-1/M1)

    International Nuclear Information System (INIS)

    Chueinta, Siripone; Julanan, Mongkol; Charncanchee, Decharchai

    2006-01-01

    The first Thai Research Reactor, TRR-1 went critical on 27 October 1962 at the maximum power of 1 MW. It was located at Office of Atoms for Peace (OAP) in Bangkok. Since then, TRR-1 was continuously operated and eventually shut down in 1975. Plate type, high-enriched uranium (HEU) and U 3 O 8 A1 cladding were used as the reactor fuel. Light water was used as moderator and coolant as well. In 1975, because of the problem from fuel supplier and also to supporting the Treaty of Non Proliferation of Nuclear Weapon or NPT, TRR-1 was shut down for modification. The reactor core and control system were disassembled and replaced by TRIGA Mark III. A new core was a hexagonal core shape designed by General Atomics (GA). Afterwards, TRR-1 was officially renamed to the Thai Research Reactor-1/Modification 1 (TRR-1/M1). TRR-1/M1 is a multipurpose swimming pool type reactor with nominal power of 2 MW. The TRR-1/M1 uses uranium enriched at 20% in U-235 (LEU) and ZrH alloy as fuel. Light water is also used as coolant and moderator. At present, the reactor is operating with core No.14. The reactor has been serving for various kinds of utilization namely, radioisotope production, neutron activation analysis, beam experiments and reactor physics experiments. (author)

  10. Argentinean integrated small reactor design and scale economy analysis of integrated reactor

    International Nuclear Information System (INIS)

    Florido, P. C.; Bergallo, J. E.; Ishida, M. V.

    2000-01-01

    This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (author)

  11. Evaluation of the trial design studies for an advanced marine reactor, (1)

    International Nuclear Information System (INIS)

    1988-03-01

    The trial design of three type reactors, semi-integrated, integrated and integrated (self-pressurized) type, was carried out in order to clarify the reactor type for the advanced marine reactor that would be developed for its realization in future and in order to extract its research and development theme. The trial design was carried and finished as for the three type reactors in same specifications in order to improve the following characteristics, small in size, light in weight, high in safety and reliability, and economic. In this report, a comparison and review of the following items are described as for the above three type reactors, (1) specifications, (2) shielding, (3) refueling, (4) in-service inspection, (5) analysis of the transients and accidents, (6) piping systems, (7) control systems, (8) dynamic analysis, (9) overall comparison, (10) research and development theme and theme for study in future. (author)

  12. Lifetime assessment on PWR reactor vessel internals in Korea

    International Nuclear Information System (INIS)

    Jung, Sung-Gyu; Jin, Tae-Eun; Jeong, Ill-Seok

    2002-01-01

    In order to extend the operating time of the Kori Unit 1 reactor internals, a comprehensive review of the potential ageing problems and a safety assessment have been performed. As the plant ages, reactor internal components which are subject to various ageing mechanism should be identified and evaluated based on the systematic technical procedure. In this respect, technical procedure for lifetime evaluation had been developed and applied to reactor internals. This paper describes a overall assessment and ageing management procedure and evaluation results for reactor internals. Also this paper suggests the optimal ageing management programs to maintain the integrity of reactor internals beyond design life based on the evaluation results. A review of all known potential ageing mechanisms was performed for each of the reactor internal subcomponents. From these results, 8 ageing mechanisms such as void swelling, irradiation and thermal embrittlement, fatigue, stress corrosion cracking, IASCC, stress relaxation, and wear for the reactor internal components were expected to be of major concerns during the current or extended plant life. In this study, 8 ageing mechanisms were identified for lifetime evaluation. For these ageing mechanisms, lifetime assessment was performed. As a result of this evaluation, it is expected that core barrel will exceed the IASCC threshold value during 40 operating years, and baffle/former and baffle former bolts will exceed the threshold value for void swelling, irradiation embrittlement, IASCC, stress relaxation during 40 operating years. However, for all other reactor internals subcomponents, thermal embrittlement, fatigue, SCC, and wear were identified as nonsignificant. As a result of lifetime evaluations, 4 ageing mechanisms were established to be plausible for 3 subcomponents. These results are shown. The existing ageing management programs (AMPs) for Kori Unit 1, such as ISI, water chemistry control, rod drop time testing etc., were

  13. Alteration of installation of nuclear reactors (alteration of No. 1 and No. 2 reactor facilities) in Shimane Nuclear Power Station, Chugoku Electric Power Co. , Inc. (report)

    Energy Technology Data Exchange (ETDEWEB)

    1987-07-01

    The Nuclear Safety Commission reported to the Minister of International Trade and Industry after prudent deliberation on this alteration problem which had been inquired on August 28, 1986. It was recognized that the technical capability of the applicant is appropriate, as the result of deliberation. It was judged that the safety after this alteration of the installation of the reactor facilities can be ensured. As the result of examining new type 8 x 8 zirconium liner fuel, there was no problem in its mechanical design, the analysis of dynamical characteristics, and the analysis of abnormal transient change and accident in operation. As to the change of the average degree of enrichment of replacement fuel, the thermonuclear design of the reactor core was adequate. In the incineration of spent resin and filter sludge, the effect of radioactive substances to the environment was negligible. The safety after abolishing the auxiliary protection function against exhaust radioactivity is ensured with a rare gas holdup equipment. The soundness of fuel and the soundness of reactor coolant pressure boundary are maintained in abnormal transient change and accident. (Kako, I.).

  14. Alteration of installation of nuclear reactors (alteration of No.1 and No.2 reactor facilities) in Shimane Nuclear Power Station, Chugoku Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1987-01-01

    The Nuclear Safety Commission reported to the Minister of International Trade and Industry after prudent deliberation on this alteration problem which had been inquired on August 28, 1986. It was recognized that the technical capability of the applicant is appropriate, as the result of deliberation. It was judged that the safety after this alteration of the installation of the reactor facilities can be ensured. As the result of examining new type 8 x 8 zirconium liner fuel, there was no problem in its mechanical design, the analysis of dynamical characteristics, and the analysis of abnormal transient change and accident in operation. As to the change of the average degree of enrichment of replacement fuel, the thermonuclear design of the reactor core was adequate. In the incineration of spent resin and filter sludge, the effect of radioactive substances to the environment was negligible. The safety after abolishing the auxiliary protection function against exhaust radioactivity is ensured with a rare gas holdup equipment. The soundness of fuel and the soundness of reactor coolant pressure boundary are maintained in abnormal transient change and accident. (Kako, I.)

  15. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

    Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

  16. Material and geometry options and performance characteristics for a test reactor

    International Nuclear Information System (INIS)

    Jahshan, S.N.; Fletcher, C.D.; Terry, W.K.

    1993-01-01

    For the past 3 yr, an Idaho National Engineering Laboratory (INEL) design team has studied design options for a new test reactor to provide continued testing services after several aging test reactors in the United States are decommissioned. This new reactor, the Broad Application Test Reactor (BATR), would also fill other currently unmet needs, such as medical isotope production and space reactor component testing. Consideration of user needs, safety requirements, developmental uncertainties, and other factors led to the selection of an evolutionary design with plate fuel and several independently cooled test loops. The fuel would be cooled by light water, but most neutron moderation would come from heavy water or beryllium. The BATR design was tentatively scaled to the Advanced Test Reactor (ATR), an existing reactor at INEL: The power output of BATR is 250 MW(thermal), and the active core heights is 1 m. For safety in loss-of-flow events, the coolant flows upward through the core. The BATR design has one large test loop (with a test space diameter of 15.0 cm) along the central axis of the core and six smaller test loops (with test space diameters of 8.0 cm) centered at 6-deg azimuthal intervals on a 24.71-cm-diam circle around the central core axis

  17. Reactor power measuring device

    International Nuclear Information System (INIS)

    Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

    1993-01-01

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

  18. Reactor container

    International Nuclear Information System (INIS)

    Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

    1997-01-01

    In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

  19. Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

    2014-01-01

    The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

  20. Project WAGR: the UK demonstration project for power reactor decommissioning - a review of the tools used to dismantle the reactor core

    International Nuclear Information System (INIS)

    Benest, T.G.

    2008-01-01

    The United Kingdom Atomic Energy Authority (UKAEA) has built and operated a wide range of nuclear facilities since the late 1940. UKAEA mission is to restore the environment of its sites in a safe and secure manner. This restoration includes the decommissioning of a number of redundant research and power reactors. The Windscale Advanced Gas-cooled Reactor (WAGR) was the UK prototype Advanced gas cooled reactor and became the forerunner of a family of 14 reactors built to generate cheaper and more efficient electricity in the UK. WAGR was constructed between 1957 and 1961 and was a carbon dioxide cooled, graphite moderated reactor using uranium oxide fuel in stainless steel cans. The reactor consisted of a graphite moderator housed in a cylindrical reactor vessel with hemispherical ends. The reactor and associated heat exchangers were enclosed in the iconic spherical containment building regularly used by the media in the UK as an illustration of the nuclear industry. The reactor first produced power in August 1962 and achieved full design output in 1963. It operated at an electrical output of 33 MW (E) for 18 years (average load factor of 75%). In 1981 the reactor was shut down after satisfactory completion of all the research and development objectives. In anticipation of the UK likely nuclear decommissioning needs the UKAEA decided to decommission WAGR to the International Atomic Energy Agency (IAEA) stage 3 (restoration of the area occupied by the facility to a condition of unrestricted re-usability) as the national demonstration exercise for power reactor decommissioning. Since 1998 the UKAEA and its contractors have been undertaking the dismantling of the reactor core components and pressure vessel in a series of 10 campaigns. These contain neutron activated components expected to produce dose rates well in excess of 1 Sv/hr. To carry out the work UKAEA installed an 8M remote dismantling machine (RDM) a waste recovery and transport system and a shielded waste

  1. Criticality design evaluation of the White Sands reactor building storage vault

    International Nuclear Information System (INIS)

    Philbin, J.S.; Nelson, W.E.

    1979-03-01

    This report describes the conceptual design and criticality evaluation of a storage vault for components of the fast pulse reactor at White Sands Missile Range. Criticality calculations were performed with the KENO-IV Monte Carlo code for various storage configurations in order to investigate the coupling between the portable reactor and storage arrays of spare reactor rings or other fissile components of similar mass. Abnormal conditions corresponding to pseudo--random arrays of the fuel components, as well as a number of flooded configurations, were also evaluated to assess criticality potential for highly unlikely situations. In a normal, dry configuration, the neutron self-multiplication factor, k/sub eff/, of the fully loaded 3 x 8 planar array plus the reactor is less than 0.87. A completely flooded vault was found to produce self-multiplication factors in excess of 1.2

  2. Nuclear power reactors: reactor safety and military and civil defence

    International Nuclear Information System (INIS)

    Hvinden, T.

    1976-01-01

    The formation of fission products and plutonium in reactors is briefly described, followed by a short general discussion of reactor safety. The interaction of reactor safety and radioactive release considerations with military and civil defence is thereafter discussed. Reactors and other nuclear plants are factors which must be taken into account in the defence of the district around the site, and as potential targets of both conventional and guerilla attacks and sabotage, requiring special defence. The radiological hazards arising from serious damage to a power reactor by conventional weapons are briefly discussed, and the benefits of underground siting evaluated. Finally the author discusses the significance of the IAEA safeguards work as a preventive factor. (JIW)

  3. Seismic analysis of a PWR 900 reactor: study of reactor building with soil-structure interaction and evaluation of floor spectra

    International Nuclear Information System (INIS)

    Gantenbein, F.; Aguilar, J.

    1983-08-01

    The purpose of this paper is the evaluation of seismic response and floor spectra for a typical PWR 900 reactor building with respect to soil-structure interaction for soil stiffness). The typical PWR 900 reactor building consists of a concrete cylindrical external building and roof dome, a concrete internal structure (internals) on a common foundation mat as illustrated. The seismic response is obtained by SRSS method and floor spectra directly from ground spectrum and modal properties of the structure. Seismic responses and floor spectra computation is performed in the case of two different ground spectra: EDF spectrum (mean of oscillator spectra obtained from 8 californian records) normalized to 0.2 g, and DSN spectrum (typical of shallow seism) normalized to 0.3 g. The first section is devoted to internals' modelisation, the second one to the axisymmetric model of the reactor, the third one to the seismic response, the fourth one to floor spectra

  4. Fuel Management of WWER-1000 Reactors of Kudankulam Nuclear Power Plant, India

    International Nuclear Information System (INIS)

    Pandey, Y.; Chauhan, A.

    2008-01-01

    Two units of WWER-1000 reactors of Russian design are under construction at Kudankulam site in India. These reactors are expected to be commissioned in 2008. The fuel management services for these reactors shall be carried out using Russian Computer codes. This paper includes a brief description of the core, fuel assembly lattice and physics modeling of the lattice and core for these reactors. Presented in this paper are the salient features of the core load pattern designs and fuel performance for 8 operating cycles of these reactors. The paper describes key improvements in the core load pattern designs to enhance the fuel utilization and its thermal behaviour. Presented in the paper are also the on site fuel management strategies with regard to fuel inventory and nuclear material accounting. A computer code for Fuel Inventory and Nuclear Material Accounting (FINMAC) has been developed for this purpose. The code FINMAC takes care of receipt of fresh fuel, flow between various accounting sub areas (ASAs), burnup or production of nuclear isotopes in the reactor cores and discharge from the reactor core. The code generates Material Balance Reports (MBRs) and Composition of Ending Inventory Reports (COEIs) as per the IAEA standards. (authors)

  5. Safeguarding on-power fuelled reactors - instrumentation and techniques

    International Nuclear Information System (INIS)

    Waligura, A.; Konnov, Y.; Smith, R.M.; Head, D.A.

    1977-05-01

    Instrumentation and techniques applicable to safeguarding reactors that are fuelled on-power, particularly the CANDU type, have been developed. A demonstration is being carried out at the Douglas Point Nuclear Generating Station in Canada. Irradiated nuclear materials in certain areas - the reactor and spent fuel storage bays - are monitored using photographic and television cameras, and seals. Item accounting is applied by counting spent-fuel bundles during transfer from the reactor to the storage bay and by placing these spent-fuel bundles in a sealed enclosure. Provision is made for inspection and verification of the bundles before sealing. The reactor's power history is recorded by a Track-Etch power monitor. Redundancy is provided so that the failure of any single piece of equipment does not invalidate the entire safeguards system. Several safeguards instruments and devices have been developed and evaluated. These include a super-8-mm surveillance camera system, a television surveillance system, a spent-fuel bundle counter, a device to detect dummy fuel bundles, a cover for enclosing a stack of spent-fuel bundles, and a seal suitable for underwater installation and ultrasonic interrogation. (author)

  6. To the analysis of reactor noise in boiling water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1972-01-01

    The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

  7. Conceptual design of ICF reactor SENRI, Part II. Advances in design and pellet gain scaling

    International Nuclear Information System (INIS)

    Ido, S.; Mima, K.; Nakai, S.; Tsuji, R.; Yamanaka, C.

    1984-01-01

    This chapter reviews the recent design studies on reactor concepts with magnetically guided lithium flow, SENRI-I, SENRI-IA and SENRI-II. The routes from the present status to power reactors and an advanced fuel pellet concept is also discussed. Topics covered include pellet design, magnetohydrodynamic design of liquid lithium flow; reactor cavity concepts with magnetically guided lithium flow, a thermo-hydraulic analysis, a tritium recovery system; and an advanced fuel pellet concept for an inertial confinement fusion (ICF) reactor without a tritium breeding blanket. An advanced fuel pellet for an ICF reactor without a T breeder was studied in the model calculations, which showed sufficiently high values of pellet gain. Includes a table and 8 diagrams

  8. The CEA research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1993-01-01

    Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

  9. High-uranium-loaded U3O8-Al fuel element development program [contributed by N.M. Martin, ORNL

    International Nuclear Information System (INIS)

    Martin, M.M.

    1993-01-01

    The High-Uranium-Loaded U 3 O 8 -Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U 3 O 8 -Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U 3 O 8 -Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U 3 O 8 ). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U 3 O 8 and aluminum. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-10

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

  11. Calorimetric measurements of the dose absorbed to graphite in the research reactors Melusine (8MWth) and Siloe (35MWth), and in the power reactor Bugey 1 (1900 MWth)

    International Nuclear Information System (INIS)

    Petitcolas, H.; Bonnin, J.J.; Chenavas, P.

    1975-01-01

    The TM calorimeter (Melusine type) developed in the CEN/Grenoble research reactors allows measurement of dose-rates over the range 10 -3 to 10 W/g. Simple and of small volume, the calorimeter causes minimum perturbation of the radiation field in which the measurement is to be made; it adapts easily to the specific requirements of the irradiation loops for which it is primarily designed. The operation is simple: measurement of temperature difference and time constants at various equilibrium positions. The results compare very favourably with those obtained by other methods, (eg. ionisation chambers) and by other workers. The calorimeter will operate for several years under irradiation and under the severe conditions of pressure, temperature etc., occurring in certain power reactors [fr

  12. Virtual nuclear reactor for education of nuclear reactor physics

    International Nuclear Information System (INIS)

    Tsuji, Masashi; Narabayashi, Takashi; Shimazu, Youichiro

    2008-01-01

    As one of projects that were programmed in the cultivation program for human resources in nuclear engineering sponsored by the Ministry of Economy, Trade and Industry, the development of a virtual reactor for education of nuclear reactor physics started in 2007. The purpose of the virtual nuclear reactor is to make nuclear reactor physics easily understood with aid of visualization. In the first year of this project, the neutron slowing down process was visualized. The data needed for visualization are provided by Monte Carlo calculations; The flights of the respective neutrons generated by nuclear fissions are traced through a reactor core until they disappear by neutron absorption or slow down to a thermal energy. With this visualization and an attached supplement textbook, it is expected that the learners can learn more clearly the physical implication of neutron slowing process that is mathematically described by the Boltzmann neutron transport equation. (author)

  13. The epithermal neutron-flux distribution in the reactor RA - Vinca

    International Nuclear Information System (INIS)

    Marinkov, V.; Bikit, I.; Martinc, R.; Veskovic, M.; Slivka, J.; Vaderna, S.

    1987-01-01

    The distribution of the epithermal neutron flux in the reactor RA - Vinca has been measured by means of Zr - activation detectors. In the channel VK-8 non-homogeneous flux distribution was observed (author) [sr

  14. Public acceptance of fusion energy and scientific feasibility of a fusion reactor. Design of inductively driven long pulse tokamak reactors: IDLT

    International Nuclear Information System (INIS)

    Ogawa, Yuichi

    1998-01-01

    Based on scientific data based adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R and D program, the scientific feasibility of inductively-driven tokamak fusion reactors is studied. A low wall-loading DEMO fusion reactor is designed, which utilizes an austenitic stainless steel in conjunction with significant data bases and operating experiences, since we have given high priority to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the DEMO reactor with the relatively large volume (i.e., major radius of 10 m) is employed, plasma ignition is achievable with a low fusion power of 0.8 GW, and an operation period of 4 - 5 hours is available only with inductive current drive. Disadvantages of pulsed operation in commercial fusion reactors include fatigue in structural materials and the necessity of an energy storage system to compensate the electric power during the dwell time. To overcome these disadvantages, a pulse length is prolonged up to about 10 hours, resulting in the remarkable reduction of the total cycle number to 10 4 during the life of the fusion plant. (author)

  15. Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

  16. Determination of the lowest critical power levels of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Binh, Do Quang; Nghiem, Huynh Ton; Tuan, Nguyen Minh; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    This paper presents the experimental methods for determining critical states of the Dalat Nuclear Research Reactor containing an extraneous neutron source induced by gamma ray reactions on beryllium in the reactor. The lowest critical power levels are measured at various moments after the reactor is shut down following 100 hours of its continuous operation. Th power levels vary from (0.5-1.2) x 10{sup -4} of P{sub n}, i.e. (25-60)W to (1.1-1.6) x 10{sup -5} of P{sub n}, i.e. (5.5-8)W at corresponding times of 4 days to 13 days after the reactor is shut down. However the critical power must be chosen greater than 500 W to sustain the steady criticality of the reactor for a long time. (author). 3 refs. 4 figs. 1 tab.

  17. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  18. Nuclear reactor

    International Nuclear Information System (INIS)

    Batheja, P.; Huber, R.; Rau, P.

    1985-01-01

    Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

  19. Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

    International Nuclear Information System (INIS)

    Reuscher, J.A.; Pickard, P.S.

    1976-01-01

    The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

  20. Identification of nuclear reactor characteristics by the reactor noise analysis

    International Nuclear Information System (INIS)

    Yashima, Hideyuki

    1980-01-01

    Reactor noise analysis method was applied to TRIGA II Research Reactor (Atomic Research Laboratory, Musashi Institute of Technology) and computed power spectral density (PSD) from the CIC current record. PSD has provided many valuable informations regarding to the reactor kinetics, including the effect of control rods vibration. Another information of neutron physics parameters were obtained and this result was compared with the parameter which was formerly measured by the Feynman-α experiment. Through these experiments we could find overall frequency characteristics of TRIGA II Reactor. (author)