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Sample records for npd2 rolphton reactor

  1. Properties of an irradiated heat-treated Zr-2.5Nb pressure tube removed from the NPD reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chow, C.K. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Koike, M.H. [Power Reactor and Nuclear Fuel Development Corp., O-Arai Engineering Centre, O-Arai (Japan); Causey, A.R.; Ells, C.E.; Hosbons, R.R.; Sagat, S.; Urbanic, V.F.; Rodgers, D.K

    1997-07-01

    Some pressure tubes in reactors moderated by heavy water have been made from heat-treated (HT) Zr-2.5Nb. One such tube was removed from the NPD nuclear reactor after 20 years of operation. An extensive program was carried out jointly by AECL and PNC to evaluate the condition and properties of this pressure tube. The investigations include irradiation creep, tensile, corrosion, delayed hydride cracking (DHC), fatigue, and fracture properties. Results show that: (I) the in-reactor elongation rate is much lower and the transverse strain rates are slightly larger than in cold-worked (CW) Zr-2.5Nb tubes; (2) the tensile properties, hydrogen pickup, threshold stress intensity factor for DHC initiation, DHC velocity, and fatigue crack growth rates were similar to those of the CW Zr-2.5Nb material; (3) the fracture toughness of this tube, as measured by curved compact toughness specimens and burst tests, is slightly higher than the CW tubes. The results were also compared with other heat-treated Zr-2.5Nb materials irradiated in the Fugen reactor. The tube was in excellent condition when removed from the reactor and would have been satisfactory for further service. (author)

  2. Fuel performance in NPD while operating with two-phase coolant

    International Nuclear Information System (INIS)

    Bain, A.S.

    1978-03-01

    The NPD reactor operated as a boiling heavy water reactor from October 27, 1968 to April 18, 1971. At 25 MWe the steam quality at the steam generator inlet was 13 wt%, and fuel channel outlet steam qualities ranged from 2 to 22 wt%. During this period ammonia was used for oxygen suppression and pH control. At equilibrium the coolant had 7 mg NH 3 /kg D 2 O, 60 ml D 2 /kg D 2 O and 20 ml N 2 /kg D 2 O. The performance of the fuel was excellent during the time that NPD operated in the boiling mode. No indications were observed of dimensional changes, inter-element fretting, fuel/sheath interaction, excessive oxidation, excessive deuterium concentrations, or unusual migration of hydrogen and deuterium to the cooler end plugs. One element defected; although the defect mechanism could not be identified at the time, we now believe the defect was associated with faulty bar stock for end plugs. The behaviour of the defective element was similar to that for other defective elements in CANDU reactors. No problems were encountered in removing the defected bundle from the reactor. (author)

  3. NPD Operating Experience

    Energy Technology Data Exchange (ETDEWEB)

    Horton, E. P. [Hydro-Electric Power Commission of Ontario, Rolphton, ON (Canada)

    1968-04-15

    NPD has demonstrated high-capacity factor operation and for the past three years has achieved an average net capacity factor of 98% for the ''winter-peak'' period. The net capacity factor for the year 1966 was 88% and for the period from the end of commissioning (October 1962) to the end of 1966 was 71%. The output of the station has been stretched from 22 MW(e) gross to 25 MW(e) gross. This was aided by the installation of an internal steam separator in the turbine but no basic modifications to the reactor-boiler systems were required. The turbine has also been modified by the installation of chrome steel diaphragms as a solution to an erosion problem. The station also continues as a test facility to develop new components and techniques. This includes the recent successful replacement of two reactor pressure tubes and the conversion of the reactor vault ventilation system to a ''dry'' atmosphere using a molecular sieve to collect heavy-water leakage and control the concentration of acidic oxides of nitrogen. Fuel performance has been excellent and the average burn-up in the core is now 84 MWh/kg U which is slightly above the equilibrium design value. Only three fuel bundles have been found with sheath failures and none of these was due to a deficiency in the fuel but was as a result of handling problems with the refuelling equipment. In spite of undesirably high maintenance time, the fuelling machines have now inserted over 1000 fuel bundles into the reactor ''on power''. Heavy-water loss rates have been acceptable and are improving. The average loss rate from leaks during 1966 was 210 g/h. A proposal to modify the NPD heavy-water heat transport system to allow boiling is under consideration. (author)

  4. Initial Operating Experience with the ''NPD'' Reactor; Experience recueillie pendant les premiers mois de fonctionnement du reacteur NPD; Pervyj opyt po ehkspluatatsii reaktora NPD; Experiencia inicial de funcionamiento del reactor NPD

    Energy Technology Data Exchange (ETDEWEB)

    McConnell, L. G. [Hydro-Electric Power Commission of Ontario, Toronto, Ontario (Canada)

    1963-10-15

    Canada's first nuclear power station, the Nuclear Power Demonstration station (NPD), is intended to serve as a means of proof-testing the performance of the Canadian type of station using natural uranium as fuel and heavy water as moderator and coolant. It reached full power on 28 June 1962. Although designed for base-load operation it will, during the early stages, be operated part of the time on high-capacity.runs and part of the time on improvement periods. Progress has been favourable so far; the first high-capacity run of six weeks'duration yielded a capacity factor of 70%. Improvements already made have increased safety, improved performance and demonstrated potential methods of capital-cost reduction for future stations. For example, shaft seals on primary coolant pumps have been modified for better performance, freezer-type vapour recovery equipment has been replaced in favour of absorption columns to reduce heavy-water vapour loss, and flow limiters are being installed in sample lines to reduce losses of heavy water in the event of joint failures. During December 1962 two simultaneous leaks from the on-power refuelling machine led to an unusual sequence of events in which a considerable amount of hot high-pressure heavy water was spilled into the reactor vault where it suffered a slight downgrading in isotopic purity. It was upgraded and the reactor returned to operation by the end of the month. All safety devices operated correctly during the incident as did the provisions for containment of heavy water. (author) [French] La premiere centrale nucleaire du Canada, NPD, est une centrale de demonstration, qui doit servir a verifier les performances des reacteurs fonctionnant a l'uranium naturel et utilisant de l'eau lourde comme ralentisseur et comme fluide de refroidissement. Elle a atteint sa pleine puissance le 28 juin 1962 bien que concue pour etre exploitee comme centrale de base, elle fonctionnera au debut comme centrale d'appoint, ce qui permettra d

  5. The NPD team conflict

    DEFF Research Database (Denmark)

    Ma, Zheng; Lin, Chih-Cheng; Tanev, Stoyan

    2012-01-01

    elaborates on the role of culture diversity and geographical dispersion in NPD team conflict. A simulation is conducted where organizations may be regarded as complex systems to affect the team conflict with a variety of influences. The results firstly indicate that there are two dimensions of NPD team...... conflict: stable and unstable dimensions with four elements: task characteristics, group members’ relationship, cultural diversity and geographical dispersion; secondly, there are two phenomena whereby the geographical dispersion influences the NPD team interaction, and the influence between cultural...

  6. New Product Development (Npd) Process In Subsidiary: Information Perspectives

    OpenAIRE

    Firmanzah

    2008-01-01

    Information is an important resource for new product development (NPD) process in subsidiary. However, we still lack of research to analyze NPD process from information perspective in subsidiary context. This research is an exploratory research and it exploited 8 cases of NPD process in consumer goods subsidiaries operating in Indonesian market. Three types of information have been identified and analyzed NPD process; global, regional and local information. The result of this research ...

  7. Lesson Learned in Preparation for Decommissioning of Three Canadian Prototype Power Reactors

    International Nuclear Information System (INIS)

    Vickerd, Meggan; Kenny, Stephen

    2016-01-01

    Lesson learned by Canadian Nuclear Laboratories (CNL)(former AECL) in preparation for decommissioning of three Prototype Reactors is a result of various strategies used for each site. CNL is responsible for the eventual decommissioning of three prototype power reactors; Nuclear Power Demonstration (NPD), Gentilly-1 and Douglas Point. Each of the Canadian prototype power reactor sites shutdown using different strategies. Depending on the site location, configuration, and intended designation of the respective sites, the individual facility systems (ventilation, electrical system, fire detection etc.) were also shut down using different strategies and operating objectives. As CNL embarks on decommissioning the first Canadian prototype reactor, this paper will reflect on the lessons learned over the past thirty years and what CNL is adjusting in the decommissioning strategy to prepare better plans for the future. The Nuclear Power Demonstration Nuclear Generating Station (NPDNGS) was constructed in late 1950's and operated from 1962 to 1987 when it was permanently shutdown after exceeding its operational goals. The NPD reactor was the first Canadian nuclear power reactor and it consisted of a single 20 MWe pressurized heavy water reactor located on a single facility site in Rolphton, Ontario. The NPD facility was shutdown to a 'Cold, Dark and Quiet' state and is maintained using an unmanned strategy by managing the site remotely with active fire detection and security surveillance systems, minimal electrical supply and an active ventilation system which is operated periodically to allow for intermittent inspections. The Douglas Point Nuclear Generating Station (DPNGS) was constructed in the early 1960's and operated from 1968 to 1984 when it was permanently shutdown. It consisted of a 200 MW prototype Canada Deuterium Uranium (CANDU) reactor and is embedded on the Bruce Power site near Kincardine, Ontario. The Douglas Point site is maintained in a

  8. Improving new product development (NPD process by analyzing failure cases

    Directory of Open Access Journals (Sweden)

    Yeon-Hak Kim

    2017-01-01

    Full Text Available Purpose - The purpose of this study is to develop an appropriate new product development (NPD process of Company “T”, a medium-sized firm, by analyzing the existing NPD process and failure cases of the Company. Design/methodology/approach - The proposed research framework is as follows: first, prospective studies of the NPD process are performed using the existing literature and preliminary references; second, comparative analysis between the current processes and a NPD process is performed; third, phase-based evaluations upon failed product cases are conducted with a NPD process so as to identify the abridged steps and root-causes of failures; finally, renewed priorities are set forth by utilizing the analytic hierarchy process analysis and questionnaire analysis upon the above identified causes of failures. Findings - The resulting accomplishments include the establishment of NPD processes that resonates with the current states of Company “T”, which, in turn, ensures the increase of efficiency, the decrease in development duration and the strategy of capacity-concentration and priority-selection. Originality/value - As Company “T”’s development process is outdated and products are developed without adequate market information research and feasibility analysis, the percentage of failed development project is as high as 87 per cent. Thus, this study aims to develop an appropriate NPD process of Company “T” by analyzing the existing NPD process and failure cases of the Company.

  9. Things get worse before getting better : outsourcing NPD activities

    NARCIS (Netherlands)

    Smets, L.P.M.; Oorschot, van K.E.; Langerak, F.

    2011-01-01

    To sustain the pace of new product development (NPD), manufacturers are increasingly outsourcing parts of the NPD process, such as design and engineering. However, by outsourcing design and engineering activities, things often get worse before getting better in terms of cycle time, development

  10. Gas and water permeability tests of 25 year old concrete from the NPD Nuclear Generating Station

    International Nuclear Information System (INIS)

    Mills, R.H.

    1990-05-01

    Permeability tests on cores recovered from concrete which had been in service for 25 years in the Nuclear Power Demonstration (NPD) reactor showed rates of mass transfer of gas and water which were greater than young fresh concrete of the same proportions and that reported in previous AECB reports. This transparency of the concrete was also 2 orders of magnitude greater than that of comparable concrete which had been stored in the laboratory atmosphere for 19 years. Analysis of the effluent in water permeability tests revealed the presence of unusual amounts of soluble materials, mainly Na and K but little Ca, in the reactor concrete. This suggested service-related deterioration of the concrete rather than the release of soluble Ca by continuing hydration of cement

  11. Information source exploitation/exploration and NPD decision-making

    DEFF Research Database (Denmark)

    Jespersen, Kristina Risom

    different Scandinavian companies. Data was analyzed using hierarchical regression models across decision criteria dimensions and NPD stages as well as analyzing the combination of selected information sources. Rather than forwarding one optimal search behavior for the entire NPD process, we find optimal...... information search behavior at either end of the exploitation/exploration continuum. Additionally, we find that overexploitation and overexploration is caused by managerial bias. This creates managerial misbehavior at gate decision-points of the NPD process.......The purpose of this study is to examine how the exploration/exploitation continuum is applied by decision-makers in new product gate decision-making. Specifically, we analyze at gate decision-points how the evaluation of a new product project is affected by the information source exploitation...

  12. When does inward technology licensing facilitate firms’ NPD performance? Evidence from China

    DEFF Research Database (Denmark)

    Li-Ying, Jason; Wang, Yuandi; Salomo, Søren

    2013-01-01

    Increasingly, many firms have found inward technology licensing a relatively quick and inexpensive way for new product development (NPD) as compared to internal R&D started. Although the literature on NPD has suggested some advantages and disadvantages regarding NPD by means of inward technology ...

  13. Is the screening of product ideas supported by the NPD process design?

    DEFF Research Database (Denmark)

    Jespersen, Kristina Risom

    2007-01-01

    literature will benefit from knowledge of the process of implementing "best practices" of NPD. Companies follow guidelines, but how is the NPD process followed through? This paper supports the need for improved insight into the complexity of screening decisions as well as knowledge of the screening...... determinants. Practical implications - The implementation of cross-functional teams and interacting phases demands more attention by management in order to reduce costs of NPD relative to the ROI of new products. Originality/value - By showing that the screening is detached from the NPD process design...

  14. PRODUCT OFFERING AS PART OF THE NPD PROCESS IN TECHNOLOGY COMPANIES

    OpenAIRE

    MATTI J. HAVERILA

    2012-01-01

    We present an exploratory investigation of how managers conceptualize and perceive the product offering effort variables in successful and unsuccessful New Product Development (NPD) projects, and explore the role that product offering effort variables play in differentiating between successful and unsuccessful NPD outcomes. Our findings indicate that managers rely on two broad types of product offering factors, "product effort" and "price/cost effort", during the NPD process so that there are...

  15. The Impact of NPD Strategy, Product Strategy, and NPD Processes on Percieved Cycle Time

    NARCIS (Netherlands)

    Parry, Mark E.; Song, Michael; Song, Michael; de Weerd-Nederhof, Petronella C.; Visscher, Klaasjan

    2009-01-01

    Studies of new product development (NPD) have identified a variety of factors that influence cycle time, but most of these findings are based on case studies of individual firms. The few empirical studies that have attempted to examine the generalizability of these findings have tended to focus on

  16. The influence of the NPD architecture on NPD decision-making in high and low technology markets

    DEFF Research Database (Denmark)

    Jespersen, Kristina Risom

    on the market. Therefore the paper investigates the influence of a stage-gate structure and cross-functional teamwork on the go/no-go decision-making within the phases of the development process for different market conditions. Data is collected with an Internet survey questionnaire from 43 large companies...... and controlling the decision-making in the NPD process thereby reducing the risk of new products failing on the market. Research finds that the structure of the new product development process and a high interaction between organizational units when developing new products enhance the likelihood of success...... on high and low technology market producing consumer products. Results reveal that the survival of product ideas within the NPD process is significant different for companies in high and low technology markets, but that the stage-gate structure of the new product development process is of less importance...

  17. The Marketplace Variables in Successful and Unsuccessful NPD Projects in Technology Intensive Companies

    OpenAIRE

    Haverila, Matti

    2010-01-01

    We present an exploratory investigation of how managers conceptualize and perceive ‘marketplace’ variables in successful and unsuccessful New Product Development (NPD) projects, and explore the role that marketplace variables play in differentiating between successful and unsuccessful NPD outcomes. Limitations and future research directions are also discussed. Our findings indicate that managers perceive the marketplace in multiple ways during the NPD process and also that differences exis...

  18. When does inward technology licensing facilitate firms' NPD performance? A contingency perspective

    DEFF Research Database (Denmark)

    Wang, Yuandi; Li-Ying, Jason

    2014-01-01

    Many firms find inward technology licensing (ITL), as a means to access external technological knowledge, an effective and relatively inexpensive way for new product development (NPD). However, although the literature has suggested some advantages and disadvantages of ITL with respect to NPD...

  19. Collaboration between competitors’ NPD teams : in search of effective modes of management control

    NARCIS (Netherlands)

    Smets, L.P.M.; Langerak, F.; Tatikonda, M.V.

    Successful collaboration between new product development (NPD) teams maintained by competitors (so-called horizontal NPD collaboration) requires the use of formal modes of management control that simultaneously coordinate the teams' interdependent development activities and encourage their

  20. Managing New Product Development Teams in a Globally Dispersed NPD Program

    DEFF Research Database (Denmark)

    Salomo, Søren; Keinschmidt, Elko J.; de Brentani, Ulrike

    2010-01-01

    Globalization is a major market trend today, one characterized by both increased international competition as well as extensive opportunities for firms to expand their operations beyond current boundaries. Effectively dealing with this important change, however, makes the management of global new...... resources, team, and performance. For the empirical analysis, data are collected through a survey of 467 corporate global new product programs (North America and Europe, business-to-business). A structural model testing for the hypothesized effects was substantially supported. The results show that creating...... product development (NPD) a major concern. To ensure success in this complex and competitive endeavor, companies must rely on global NPD teams that make use of the talents and knowledge available in different parts of the global organization. Thus, cohesive and well-functioning global NPD teams become...

  1. Supporting NPD teams in innovation : Structuring user data on the foundations of empathy

    NARCIS (Netherlands)

    Alkaya, M.; Sleeswijk Visser, F.; de Lille, C.S.H.

    2012-01-01

    Gathering and analyzing user data in new product development (NPD), both qualitative and quantitative, can be a strong driver for innovation in companies. While ideally user data should be obtained through direct interactions between NPD departments and targeted users, time and cost restrictions

  2. The Marketplace Variables in Successful and Unsuccessful NPD Projects in Technology Intensive Companies

    Directory of Open Access Journals (Sweden)

    Matti J. Haverila

    2010-12-01

    Our findings indicate that managers perceive the marketplace in multiple ways during the NPD process and also that differences exist in metric equivalence across successful and unsuccessful NPD projects. Also, although half of the marketplace variables are positively related to NPD success, managers in Finnish technology companies appear to attach higher relative importance to market attractiveness rather than market competitiveness variables. Marketplace variables appear to be less important than in the Korean and Chinese samples, and much more important than  in the Canadian sample in the Mishra et all study (1996, and similarly much more important than in the Cooper study (1979b.

  3. NPD gate decision criteria - a consequence of strategic orientation or decision-maker expertise

    DEFF Research Database (Denmark)

    Søndergård, Helle Alsted Søndergård; Jespersen, Kristina Risom

    2008-01-01

    criteria at the decision gates. Results from a sample of 132 NPD practitioners studied a simulated NPD process show that the importance of decision criteria is influenced by both the strategic orientation of the company and individual decision-making expertise. Especially the proactive and aggressive...... strategic orientations influence the prioritisation of decision criteria....

  4. New Product Development (NPD) Process - An Example of Industrial Sector

    Science.gov (United States)

    Kazimierska, Marianna; Grębosz-Krawczyk, Magdalena

    2017-12-01

    This aim of this article is to present the process of new product introduction on example of industrial sector in context of new product development (NPD) concept. In the article, the concept of new product development is discussed and the different stages of the process of new electric motor development are analysed taking into account its objectives, implemented procedures, functions and responsibilities division. In the article, information from secondary sources and the results of empirical research - conducted in an international manufacturing company - are used. The research results show the significance of project leader and regular cooperation with final client in the NPD process.

  5. Luncheon address: Development of the CANDU reactor

    International Nuclear Information System (INIS)

    Bain, A.S.

    1997-01-01

    The paper is a highlight of the some of the achievements in the development of the CANDU Reactor, taken from the book C anada Enters the Nuclear Age . The CANDU reactor is one of Canada's greatest scientific/engineering achievements, that started in the 1940's and bore fruit with the reactors of the 60's, 70's, and 80's. The Government decided in the 1950's to proceed with a demonstration nuclear power reactor (NPD), AECL invited 7 Canadian corporations to bid on a contract to design and construct the NPD plant. General Electric was selected. A utility was also essential for participation and Ontario Hydro was chosen. In May 1957 it was concluded that the minimum commercial size would be about 200MWe and it should use horizontal pressure tubes to contain the fuel and pressurized heavy water coolant. The book also talks of standard out-reactor components such as pumps, valves, steam generators and piping. A major in-reactor component of interest was the fuel, fuel channels and pressure tubes. A very high level of cooperation was required for the success of the CANDU program

  6. Launch Effort and NPD Success: A Study of Technology Intensive Companies in Finland

    OpenAIRE

    Matti J. Haverila; Nicholas Ashill

    2014-01-01

    In this paper, we present an exploratory investigation of the types of product launch tactics managers use in technology-intensive companies in Finland and explore the role that such tactics play in differentiating between successful and unsuccessful NPD outcomes. Our findings indicate that managers perceive two types of tactical product launch tactics during the NPD process, which we label "promotions and sales-distribution effort" and "production start-up effort". All tactical launch effort...

  7. Fuel deposits, chemistry and CANDU® reactor operation

    International Nuclear Information System (INIS)

    Roberts, J.G.

    2014-01-01

    'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU® reactor, the first being the Nuclear Power Demonstration - 2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channelled to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5. The difference being that during 'hot conditioning' of CANDU® heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

  8. Postmortem identification and quantitation of 2,5-dimethoxy-4-n-propylthiophenethylamine using GC-MSD and GC-NPD.

    Science.gov (United States)

    Curtis, Byron; Kemp, Philip; Harty, Linda; Choi, Chai; Christensen, Dix

    2003-10-01

    2,5-Dimethoxy-4-n-propylthiophenethylamine (2C-T-7) has structural and pharmacodynamic similarities to methylenedioxymethamphetamine (MDMA). This compound was initially identified from a routine screening procedure in postmortem urine from a 20-year-old male that died in a local emergency room after reportedly insufflating 35 mg. This report describes the development of a quantitative method for 2C-T-7. A number of method parameters were studied including internal standard selection, liquid-liquid extraction scheme, and drug stability in preserved refrigerated blood. The adopted method for blood and urine involves the addition of trimethoxyamphetamine (TMA) as internal standard, alkalinization with ammonium hydroxide, and liquid-liquid extraction with n-chlorobutane. To facilitate recovery from liver, a 1:4 aqueous homogenate was pretreated with dilute perchloric acid, centrifuged, and the supernatant was extracted as previously described. In each case, 0.1% hydrochloric acid in methanol was added during the final concentration step to prevent loss of drug caused by evaporation. Samples were analyzed by gas chromatography with nitrogen-phosphorus detection (GC-NPD) and electron ionization GC-mass spectrometry (MS) utilizing selected ion monitoring. For the GC-MS analysis, the characteristic ions monitored for 2C-T-7 were m/z 226, 255, and 183 and for TMA, m/z 182. The limits of detection and quantitation in blood were 6.0 and 15.6 ng/mL, respectively, by both GC-NPD and GC-MS. The results from the postmortem case were as follows: heart blood, 57 ng/mL; femoral blood, 100 ng/mL; urine, 1120 ng/mL; and liver, 854 ng/g.

  9. Improvement of CSCW Software Implementation in NPD: The CAM Mechanism for a Better Adoption by Users

    Science.gov (United States)

    Restrepo, Tomas; Arbelaez, Natalia; Millet, Dominique; Gidel, Thierry

    2010-01-01

    Cooperation between disseminated actors is a key factor in improving new product development (NPD) performance. In the last years, numerous CSCW software applications have been introduced in the industry to support NPD with a low success rate. This is partly due to the limited insight of the organisational and human factors influencing user…

  10. Virtual Competencies and Knowledge Transfer in Global NPD: A Case Study

    Directory of Open Access Journals (Sweden)

    Päivi Lohikoski

    2013-12-01

    Full Text Available Virtual new product development teams are geographically dispersed and cross-functional, yet they work on highly interdependent tasks by communicating electronically in work groups. This virtual interdependence, among other issues, presents new challenges for the management of knowledge transfer in global New Product Development (NPD. In this project, virtual competencies were studied with the use of qualitative methods to assess the most significant issues affecting knowledge transfer in virtual settings in global NPD. As a result, the existing theory regarding virtual competencies and virtual organizations was refined and potential barriers for knowledge transfer were discovered. The success of knowledge workers is crucial for the performance of knowledge-based organizations, which form the basis of our global economy; therefore, this study’s findings are significant.

  11. Escalation, de-escalation, or reformulation : effective interventions in delayed NPD projects

    NARCIS (Netherlands)

    Oorschot, van K.E.; Langerak, F.; Sengupta, K.

    2011-01-01

    At the start of any project, new product development (NPD) teams must make accurate decisions about development time, development costs, and product performance. Such profit-maximizing decision making becomes particularly difficult when the project runs behind schedule and requires intervention

  12. 4P-NPD ultra-thin films as efficient exciton blocking layers in DBP/C70 based organic solar cells

    Science.gov (United States)

    Patil, Bhushan R.; Liu, Yiming; Qamar, Talha; Rubahn, Horst-Günter; Madsen, Morten

    2017-09-01

    Exciton blocking effects from ultra-thin layers of N,N‧-di-1-naphthalenyl-N,N‧-diphenyl [1,1‧:4‧,1″:4″,1‴-quaterphenyl]-4,4‴-diamine (4P-NPD) were investigated in small molecule-based inverted organic solar cells (OSCs) using tetraphenyldibenzoperiflanthene as the electron donor material and fullerene (C70) as the electron acceptor material. The short-circuit current density (J SC) and power conversion efficiency (PCE) of the optimized OSCs with 0.7 nm thick 4P-NPD were approximately 16% and 24% higher, respectively, compared to reference devices without exciton blocking layers (EBLs). Drift diffusion-based device modeling was conducted to model the full current density-voltage (JV) characteristics and external quantum efficiency spectrum of the OSCs, and photoluminescence measurements were conducted to investigate the exciton blocking effects with increasing thicknesses of the 4P-NPD layer. Importantly, coupled optical and electrical modeling studies of the device behaviors and exciton generation rates and densities in the active layer for different 4P-NPD layer thicknesses were conducted, in order to gain a complete understanding of the observed increase in PCE for 4P-NPD layer thicknesses up to 1 nm, and the observed decrease in PCE for layer thicknesses beyond 1 nm. This work demonstrates a route for guiding the integration of EBLs in OSC devices.

  13. Fuel deposits, chemistry and CANDU reactor operation

    International Nuclear Information System (INIS)

    Roberts, J.G.

    2013-01-01

    'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU reactor, the first being the Nuclear Power Demonstration-2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channel led to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5, and subsequently utilized for each CANDU unit since. The difference being that during 'hot conditioning' of CANDU heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

  14. Improving sourcing decisions in NPD projects: Monetary quantification of points of difference

    NARCIS (Netherlands)

    Wouters, Marc; Anderson, James C.; Narus, James A.; Wynstra, Finn

    2009-01-01

    During new product development (NPD), firms make critical design and sourcing decisions that determine the new product's cost, performance, competitive position, and profitability. The purchase price of materials and components for the new product provides only part of the picture for design and

  15. Enhancing New Product Development (NPD Portfolio Performance by Shaping the Development Funnel

    Directory of Open Access Journals (Sweden)

    Paulo Soares Figueiredo

    2012-12-01

    Full Text Available New product development (NPD projects are typically managed through a series of screens, or gates, where ideas compete for resources. Ideas are carved into projects, and these projects are reviewed, and approved or terminated through the screening process so that only the best performing projects continue to subsequent stages of design, development and testing, and are released into the market place (Krishnan and Ulrich 2001; Terwiesch and Ulrich 2009. Most large innovative organizations deal with more than one NPD project at a time and typically engage in product pipeline management (PPM, where a set of active projects are evaluated together while they traverse through a sequence of such screens. Key decisions in a R&D pipeline are: screen thresholds, complexity of projects, resource allocation and capacity adjustment biases. We explore the impact of structural and behavioral aspects of these decisions through a simulation based analysis of a pharmaceutical dataset. Results establish concave relationships between value created at the end of pipeline and the resource allocation and complexity allocation biases, indicating optimizability and a limit for front loading practices.

  16. The impact of team cognitive styles on performance of radical and incremental NPD projects

    NARCIS (Netherlands)

    de Visser, Matthias; Faems, Dries; Visscher, Klaasjan; de Weerd-Nederhof, Petronella C.

    2014-01-01

    Although prior studies increased our understanding of the performance implications of new product development (NPD) team members’ functional backgrounds and demographic variables, they remained relatively silent on the impact of underlying psychological characteristics such as the team members’

  17. The Impact of Team Cognitive Styles on Performance of Radical and Incremental NPD Projects

    NARCIS (Netherlands)

    de Visser, Matthias; Faems, Dries; Visscher, Klaasjan; de Weerd-Nederhof, Petra

    2014-01-01

    Although prior studies increased our understanding of the performance implications of new product development (NPD) team members' functional backgrounds and demographic variables, they remained relatively silent on the impact of underlying psychological characteristics such as the team members'

  18. The cluster burn up programme CCC and a comparison of its results with NPD experiments

    International Nuclear Information System (INIS)

    Hoejerup, C.F.

    1976-10-01

    A brief description is given of the computer programme CCC, which can be used for rod/rod cluster burn up calculations. A comparison of CCC results with some Canadian measurements on NPD fuel is also included. (author)

  19. 4P-NPD ultra thin-films as efficient exciton blocking layers in DBP/C70 based organic solar cells

    DEFF Research Database (Denmark)

    Patil, Bhushan Ramesh; Liu, Yiming; Qamar, Talha

    2017-01-01

    Exciton blocking effects from ultra thin layers of N,N'-di-1-naphthalenyl-N,N'-diphenyl [1,1':4',1'':4'',1'''-quaterphenyl]-4,4'''-diamine (4P-NPD) was investigated in small molecule based inverted Organic Solar Cells (OSCs) using Tetraphenyldibenzoperiflanthene (DBP) as the electron donor material...... and fullerene (C70) as the electron acceptor material. The short-circuit current density (Jsc) and PCE of the optimized OSCs with 0.7 nm thick 4P-NPD were approx. 16 % and 24 % higher, respectively, compared to reference devices without exciton blocking layers. Drift diffusion based device modeling...... was conducted to model the full Current density – Voltage (JV) characteristics and EQE spectrum of the OSCs, and photoluminescence measurements was conducted to investigate the exciton blocking effects with increasing thicknesses of the 4P-NPD layer. Importantly, coupled optical and electrical modeling studies...

  20. Exploring the link between strategy and decision criteria in NPD

    DEFF Research Database (Denmark)

    Søndergaard, Helle Alsted; Jespersen, Kristina Risom

    process. Respondents with a prospector strategy use DC to a higher degree than analysers and respondents with the highest level of innovativeness use soft DC to a higher degree compared to those with low innovativeness who use financial criteria at all gates. The relationship between strategy and newness......Our major objective is to investigate the link between strategy and decision criteria (DC). An exploratory study was conducted by means of questionnaire-data and in-depth interviews with respondents involved in NPD. Results show extensive use of strategic DC, primarily early in the development...

  1. Evolution of CANDU reactor design

    International Nuclear Information System (INIS)

    Pon, G.A.

    1978-08-01

    The CANDU (CANada Deuterium Uranium) design had its begin-ings in the early 1950's with the preliminary engineering studies that led to the 20 MW(e) NPD (Nuclear Power Demonstration) and the 200 MW(e) Douglas Point station . The next decade saw the first operation of both these stations and the commitment of the 2000 MW(e) Pickering and 3000 MW(e) Bruce plants. The present decade has witnessed the excellent performance of Pickering and Bruce and commitments to construct Gentilly-2, Cordoba, Pt. Lepreau, Wolsung, Pickering B, Bruce B and Darlington. In most cases, successive CANDU designs have meant an increase in plant output. Evolutionary developments have been made to fit the requirements of higher ratings and sizes, new regulations, better reliability and maintainability and lower costs. These changes, which are described system by system, have been introduced in the course of engineering parallel reactor projects with overlapping construction schedules -circumstances which ensure close contact with the practical realities of economics, manufacturing functions, construction activities and performance in commissioning. Features for one project furnished alternative concepts for others still on the drawing board and the experience gained in the first application yielded a sound basis for its re-use in succeeding projects. Thus the experiences gained in NPD, Douglas Point, Gentilly-1 and KANUPP have contributed to Pickering and Bruce, which in turn have contributed to the design of Gentilly-2. (author)

  2. Small power reactor projects in the United States of America and Canada. Information gathered as a result of invitations from Member States

    International Nuclear Information System (INIS)

    1962-01-01

    As part of its activities in connection with the development of nuclear power, and in response to the resolutions adopted by the General Conference, the Agency has been undertaking a continuing study of the technology and economics of small and medium sized power reactors, particularly with reference to the needs of the less-developed countries. This report summarizes the information gathered on the small power reactor projects in the United States of America and Canada, as a result of the opportunity afforded by these Member States to the Agency. It may be recalled that, at the third regular session of the General Conference, the United States Government offered to provide the Agency with relevant technical and economic data on several small power reactor projects of its Atomic Energy Commission. The Agency accepted the offer and since June 1960 it has sent one or two staff members at approximately six-monthly intervals to follow the development of nine power reactor projects in the United States which represent six different reactor systems. Last year, the Agency issued a report summarizing the information obtained through their visits and study of available published literature. The present document, which should be read in conjunction with that document, brings the information up to date and provides additional information on certain phases of the projects already discussed in the last report. Three more power reactor projects are also dealt with, namely the experimental gas-cooled reactor (EGCR), the high temperature gas-cooled reactor (HTGR) and the Hallam nuclear power facility (HNPF). Early in 1962, the Canadian Government expressed its willingness to make available to the Agency relevant information on the NPD and CANDU projects. The coverage of the NPD reactor is based upon the published information supplied by AECL of Canada and the visit by one of the staff members to the NPD site. The Agency wishes to acknowledge with thanks the co-operation extended

  3. Influence of Pentacene Interface Layer in ITO/α-NPD/Alq3/Al Organic Light Emitting Diodes by Time-Resolved Electric-Field-Induced Optical Second-Harmonic Generation Measurement.

    Science.gov (United States)

    Oda, Yoshiaki; Sadakata, Atsuo; Taguchi, Dai; Manaka, Takaaki; Iwamoto, Mitsumasa

    2016-04-01

    By using I-V, EL-V, displacement current measurement (DCM) and time-resolved electric-field-induced optical second-harmonic generation (TR-EFISHG) measurement, we studied the influence of interface pentacene layer inserted between ITO and a-NPD layers in ITO/α-NPD/Alq3/Al OLEDs. All experiments were carried out for the OLEDs with and without a pentacene interface layer. The I-V and EL-V measurements showed the decrease of operating voltage of EL, the DCM showed the lowering of inception voltage of carrier injection by inserting a pentacene interface layer. The TR-EFISHG measurement showed the faster accumulation of holes at the interface between the a-NPD and Alq3 layers, which resulted in the relaxation of electric field of a-NPD layer accomplished by the increase of the conductivity and the increase of the electric field in the Alq3 layer. We conclude that TR-EFISHG measurement is helpful for understanding I-V and EL-V characteristics, and can be combined with other methods to give significant information which are impacted by the interface layer.

  4. Interactions in New Product Development: how the nature of the NPD process influences interaction between teams and management

    DEFF Research Database (Denmark)

    Felekoglu, Burcu; Maier, Anja; Moultrie, James

    2013-01-01

    Effective interaction across organisational boundaries is a critical success factor in new product development (NPD). However, few studies have investigated how different mechanisms enable effective interaction across organisational and particularly hierarchical boundaries. This study explores how...

  5. MACS, the manipulation and collimation system on the NPD at LANSCE

    International Nuclear Information System (INIS)

    Bourke, M.A.M.; Goldstone, J.A.; Lovell, K.J.

    1991-01-01

    The practical problems associated with beam collimation and specimen manipulation at a pulsed neutron source are identical to those on a steady state source. However extra constraints result from the limited space available and from the time of flight analysis of the diffracted neutrons. A manipulation and collimation system (MACS) has been designed for the neutron powder diffractometer (NPD) at the Los Alamos spallation neutron source (LANSCE). It provides specimen motion and aperture positioning with accuracies of better than 0.1mm and is constructed as a rigid unit. For flat sided specimens sampling volumes less than 30 mm 3 have been obtained demonstrating the viability of making spatially resolved strain measurements at a pulsed neutron source. 3 figs

  6. Selected power reactor projects in Canada and the United States of America

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-11-01

    As part of its activities in connection with the development of nuclear power, the IAEA has undertaken a continuing study of the technology and economics of power reactors, with particular reference to the needs of the developing countries. Information on the progress made in eight power reactor projects, namely those of Bonus, Pathfinder, Elk River, Piqua, Hallam, Experimental Gas-Cooled Reactor (EGCR), High-Temperature Gas-Cooled Reactor (HTGCR) and Nuclear Power Demonstration (NPD), is presented in this report. Developments during the past year are shown, emphasis being placed on operating experience in the case of those reactors which have become critical. The Agency is grateful to the Governments of Canada and the USA, who have extended the necessary facilities for covering he different power reactor projects in their respective countries. The cooperation received from the reactor manufacturers, builders and operators is also gratefully acknowledged. It is hoped that this report will be of interest to reactor technologists and operators and those interested in the application of nuclear power.

  7. Selected power reactor projects in Canada and the United States of America

    International Nuclear Information System (INIS)

    1964-01-01

    As part of its activities in connection with the development of nuclear power, the IAEA has undertaken a continuing study of the technology and economics of power reactors, with particular reference to the needs of the developing countries. Information on the progress made in eight power reactor projects, namely those of Bonus, Pathfinder, Elk River, Piqua, Hallam, Experimental Gas-Cooled Reactor (EGCR), High-Temperature Gas-Cooled Reactor (HTGCR) and Nuclear Power Demonstration (NPD), is presented in this report. Developments during the past year are shown, emphasis being placed on operating experience in the case of those reactors which have become critical. The Agency is grateful to the Governments of Canada and the USA, who have extended the necessary facilities for covering he different power reactor projects in their respective countries. The cooperation received from the reactor manufacturers, builders and operators is also gratefully acknowledged. It is hoped that this report will be of interest to reactor technologists and operators and those interested in the application of nuclear power

  8. Structural ambidexterity in NPD processes; A firm-level assessment of the impact of differentiated structures on innovation performance

    NARCIS (Netherlands)

    de Visser, Matthias; de Weerd-Nederhof, Petronella C.; Faems, D.L.M.; Song, Michael; Song, Michael; van Looy, Bart; van Looy, Bart; Visscher, Klaasjan

    2010-01-01

    Based on a survey study of 155 U.S. firms, we conducted a firm-level assessment of the impact of different kinds of structures (i.e., functional versus cross-functional) in different kinds of new product development (NPD) processes (i.e., incremental versus radical) on different kinds of firm

  9. STRATEGI NEW PRODUCT DEVELOPMENT (NPD-PROCESS PADA PERUSAHAAN MANUFAKTUR YANG BERBASIS INOVASI

    Directory of Open Access Journals (Sweden)

    Yusaq Tomo Ardianto

    2009-02-01

    Full Text Available Pemberlakuan perdagangan bebas ACFTA menjadikan Asia Tenggara sebagai potensi pasar yang sangat besar, selain itu juga sebagai potensi produksi yang juga cukup besar, sehingga banyak pengusaha ataupun investor untuk menanamkan modalnya di wilayah yang sedang tumbuh dan berkembang ini. Kunci sukses perusahaan di dalam memenangkan persaingan salah satunya adalah dengan cara melakukan kegiatan usahanya dekat dengan pasar dan mengembangkan/melakukan inovasi produk baru atau NPD.  Analisis konseptual ini khususnya memberikan penjelasan tentang sebuah produk bukan jasa, bagaimana produk baru tersebut berproses dan berkembang, dan bagaimana inovasi yang dilakukan oleh perusahaan manufaktur pada produk yang dikembangkannya dan alasan-alasan mengapa perlu dilakukannya pengembangan produk baru dan inovasi produk

  10. Reactor BR2

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2000-07-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

  11. Reactor BR2

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

  12. Evolution of on-power fuelling machines on Canadian natural uranium power reactors

    International Nuclear Information System (INIS)

    Isaac, P.

    1984-10-01

    The evolution of the on-power fuel changing process and fuelling machines on CANDU heavy-water pressure tube power reactors from the first nuclear power demonstration plant, 22 MWe NPD, to the latest plants now in design and development is described. The high availability of CANDU's is largely dependent on on-power fuelling. The on-power fuelling performance record of the 16 operating CANDU reactors, covering a 22 year period since the first plant became operational, is given. This shows that on-power fuel changing with light (unshielded), highly mobile and readily maintainable fuelling machines has been a success. The fuelling machines have contributed very little to the incapabilities of the plants and have been a key factor in placing CANDUs in the top ten list of world performance. Although fuel handling technology has reached a degree of maturity, refinements are continuing. A new single-ended fuel changing concept for horizontal reactors under development is described. This has the potential for reducing capital and operating costs for small reactors and increasing the fuelling capability of possible large reactors of the future

  13. Constraints and Creativity in NPD - Testing the Impact of 'Late Constraints'

    DEFF Research Database (Denmark)

    Onarheim, Balder; Valgeirsdóttir, Dagný

    experiment was conducted, involving 12 teams of industrial designers from three different countries, each team working on two 30 minutes design tasks. In one condition all constraints were given at the start, and in the other one new radical constraint was added after 12 minutes. The output from all 24 tasks......The aim of the presented work is to investigate how the timing of project constraints can influence the creativity of the output in New Product Development (NPD) projects. When seeking to produce a creative output, is it beneficial to know all constraints when initiating a project...... was assessed for creativity using the Consensual Assessment Technique (CAT), and a comparative within-subjects analysis found no significant different between the two conditions. Controlling for task and assessor a small but non-significant effect was found, in favor of the ‘late constraint’ condition. Thus...

  14. PERIGEE computer codes for reactor simulation in 3 dimensions, using 1 or 2 neutron velocity groups

    International Nuclear Information System (INIS)

    Olson, A.P.

    1964-02-01

    PERIGEE is a code written in SNAP for the G-20 computer. It solves the one- or two-group neutron diffusion equations by finite-difference methods on a three-dimensional, uniform mesh having a common spacing in the two directions normal to the fuel channels. The positions of mesh points along a fuel channel, relative to points in adjacent channels, may correspond to either NPD or CANDU fuel bundle positions. The extrapolated flux boundary may be specified in sufficient detail to represent a tapered or stepped circumferential reflector, a variable axial length and, for a reactor with axis horizontal, a variable moderator level and a variable plane bottom surface equivalent to the CANDU dump structure. The neutron flux may be normalized to give a specified power output from the hottest fuel bundle or hottest channel, or to give a total thermal power limited by the turbine and generator. Reactor operation may be simulated in finite time steps, taking into account any fuel shifts, any changes in moderator level and the change in nuclear properties of the fuel with increasing irradiation. The appropriate properties are obtained by interpolation from tables supplied for as many as 8 types of fuel bundle. The mean fuel exit burnup can be calculated at equilibrium for a reactor in which the exit burnups for two zones may be adjusted to give radial power flattening and the fuelling schedules may be designed to give axial power flattening in one or both zones. (author)

  15. The Use of Absorptive Capacity in Improving the New Product Development (NPD

    Science.gov (United States)

    Gunawan, W.; Gerardus, P.; Tji, B. J.; Richard, K.

    2017-01-01

    The term Absorptive Capacity (AC) refers to maximizing the external knowledge transfer into the organization in order to improve its performance. Since its introduction in year 1990, AC has been applied widely in many fields such as: economy, business, KM, HR, intellectual capital, IT, operation management, marketing, etc. Due to its wide application, nevertheless, The AC application in both Indonesian industry and higher education institutions (HEIs) are still rare to find. The Indonesian Directorate General of Higher Education (DGHE) has encouraged creating effective collaboration model that enables to link the HEIs with the industries in order to improve knowledge creation in improving product development that can be used by the firms. For this reason, the article examines the effective AC model that enables to assist in improving new product development (NPD) process in the polytechnic perspectives.

  16. Summary of feasibility studies on in situ disposal as a decommissioning option for nuclear facilities

    International Nuclear Information System (INIS)

    Helbrecht, R.A.

    2002-01-01

    A scoping study was conducted over the period 1998-2000 to consider the feasibility of in situ disposal as a decommissioning option for AECL's Nuclear Power Demonstration Reactor located at Rolphton, Ontario. The results of a detailed assessment are summarized and the study concludes that in situ disposal appears feasible. Additional work required to confirm the results is also identified. A second in situ component, contaminated Winnipeg River sediments at AECL's Whiteshell Laboratory located in Manitoba, was also evaluated. That study concluded that in situ abandonment would have no adverse impact on aquatic life, humans and the environment. A summary of the study is presented as an appendix to the report. (author)

  17. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

    Wyne, M.F.; Meghji, J.H.

    1990-12-01

    PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

  18. Reactor BR2: Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. A safety audit was conduced by the IAEA, the conclusions of which demonstrated the excellent performance of the plant in terms of operational safety. In 1999, the CALLISTO facility was extensively used for various programmes involving LWR pressure vessel materials, IASCC of LWR structural materials, fusion reactor materials and martensic steels for use in ADS systems. In 1999, BR2's commercial programmes were further developed

  19. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  20. EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

    International Nuclear Information System (INIS)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  1. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2001-01-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

  2. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  3. Safety features of TR-2 reactor

    International Nuclear Information System (INIS)

    Tuerker, T.

    2001-01-01

    TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

  4. Sources of innovation, their combinations and strengths – benefits at the NPD project level

    DEFF Research Database (Denmark)

    Tranekjer, Tina Lundø; Søndergaard, Helle Alsted

    2013-01-01

    External sourcing is increasingly seen as important for obtaining new and valuable knowledge and resources for new product development. However, when it comes to the specifics of choosing between sources and types of relationships, little is known on the NPD project level. This paper strengthens...... not only consider the potential benefits of collaboration with external sources but also the downsides, including higher cost and lengthier projects. Firms should look for opportunities in the combination of sources if they are to gain advantages of collaboration, as our analyses show that a mix of market...... and science sources is related to decreased costs. Additionally, if firms are looking for increased market performance, they should aim at collaborating with suppliers that have a similar knowledge base, whereas if the aim is lower project costs, collaboration with a customer with a similar knowledge base...

  5. Determination of prilocaine HCl in bulk drug and pharmaceutical formulation by GC-NPD method

    Directory of Open Access Journals (Sweden)

    Atila Alptug

    2013-01-01

    Full Text Available The novel analytical method was developed and validated for determination of prilocaine HCl in bulk drug and pharmaceutical formulation by gas chromatography-nitrogen phosphorus detection (GC-NPD. The chromatographic separation was performed using a HP-5MS column. The calibration curve was linear over the concentration range of 40-1000 ng ml-1 with a correlation coefficient of 0.9998. The limits of detection (LOD and quantification (LOQ of method were 10 ng ml-1 and 35 ng ml-1, respectively. The within-day and between-day precision, expressed as the percent relative standard deviation (RSD% was less than 5.0%, and accuracy (percent relative error was better than 4.0%. The developed method can be directly and easily applied for determination of prilocaine HCl in bulk drug and pharmaceutical formulation using internal standard methodology.

  6. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2002-01-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system

  7. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2002-04-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

  8. Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, P. C. [Canadian General Electric (Canada)

    1968-07-15

    The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

  9. BR2 Reactor: Introduction

    International Nuclear Information System (INIS)

    Moons, F.

    2007-01-01

    The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3 He, 6 Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

  10. The effect of innovation competence on the choice of information elements in a simulated NPD process

    DEFF Research Database (Denmark)

    Søndergaard, Helle Alsted; Juhl, Hans Jørn

    to prompt different managerial behaviour. This study uses a role-play approach to investigating the relationship between innovation competence (exploitation or exploration competences) and the acquisition of information elements in a simulated new product development (NPD) process. Results show that two......This paper presents a study of the relationship between innovation competence and managerial behaviour. Within the resource-based view of the firm the development of new products is often related to the attainment of dynamic capabilities. Different levels of dynamic capabilities are expected...... distinct clusters can be identified among respondents, one cluster with both exploitation and exploration competence experience and a second cluster only with exploitation competence experience. Contrary to expectations analysis of variance and Chi-square tests for independence show that there are no clear...

  11. Application of MCNPX 2.7.D for reactor core management at the research reactor BR2

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, Edgar

    2011-01-01

    The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK-CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full core model. The second method represents fully automatic whole core depletion and criticality calculations in the detailed 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared vs. the developed at the BR2 reactor department MCNPX & ORIGEN-S combined method. Testing of MCNPX 2.7.D on the criticality measurements at the BR2 reactor is presented. (author)

  12. 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

  13. 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

  14. Prevalence, correlates, disability, and comorbidity of DSM-IV narcissistic personality disorder: results from the wave 2 national epidemiologic survey on alcohol and related conditions.

    Science.gov (United States)

    Stinson, Frederick S; Dawson, Deborah A; Goldstein, Risë B; Chou, S Patricia; Huang, Boji; Smith, Sharon M; Ruan, W June; Pulay, Attila J; Saha, Tulshi D; Pickering, Roger P; Grant, Bridget F

    2008-07-01

    To present nationally representative findings on prevalence, sociodemographic correlates, disability, and comorbidity of narcissistic personality disorder (NPD) among men and women. Face-to-face interviews with 34,653 adults participating in the Wave 2 National Epidemiologic Survey on Alcohol and Related Conditions conducted between 2004 and 2005 in the United States. Prevalence of lifetime NPD was 6.2%, with rates greater for men (7.7%) than for women (4.8%). NPD was significantly more prevalent among black men and women and Hispanic women, younger adults, and separated/divorced/widowed and never married adults. NPD was associated with mental disability among men but not women. High co-occurrence rates of substance use, mood, and anxiety disorders and other personality disorders were observed. With additional comorbidity controlled for, associations with bipolar I disorder, post-traumatic stress disorder, and schizotypal and borderline personality disorders remained significant, but weakened, among men and women. Similar associations were observed between NPD and specific phobia, generalized anxiety disorder, and bipolar II disorder among women and between NPD and alcohol abuse, alcohol dependence, drug dependence, and histrionic and obsessive-compulsive personality disorders among men. Dysthymic disorder was significantly and negatively associated with NPD. NPD is a prevalent personality disorder in the general U.S. population and is associated with considerable disability among men, whose rates exceed those of women. NPD may not be as stable as previously recognized or described in the DSM-IV. The results highlight the need for further research from numerous perspectives to identify the unique and common genetic and environmental factors underlying the disorder-specific associations with NPD observed in this study.

  15. Core monitoring at the WNP-2 reactor

    International Nuclear Information System (INIS)

    Skeen, D.R.; Torres, R.H.; Burke, W.J.; Jenkins, I.; Jones, S.W.

    1992-01-01

    The WNP-2 reactor is a 3,323-MW(thermal) boiling water reactor (BWR) that is operated by the Washington Public Power Supply System. The WNP-2 reactor began commercial operation in 1984 and is currently in its eighth cycle. The core monitoring system used for the first cycle of operation was supplied by the reactor vendor. Cycles 2 through 6 were monitored with the POWERPLEX Core Monitoring Software System (CMSS) using the XTGBWR simulation code. In 1991, the supply system upgraded the core monitoring system by installing the POWERPLEX 2 CMSS prior to the seventh cycle of operation for WNP-2. The POWERPLEX 2 CMSS was developed by Siemens Power Corporation (SPC) and is based on SPC's advanced state-of-the-art reactor simulator code MICROBURN-B. The improvements in the POWERPLEX 2 system are possible as a result of advances in minicomputer hardware

  16. TA-2 Water Boiler Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Durbin, M.E.; Montoya, G.M.

    1991-06-01

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

  17. An Efficient Model for NPD Performance Evaluation Using DEMATEL and Fuzzy ANP—Applied to the TFT-LCD Touch Panel Industry in Taiwan

    Directory of Open Access Journals (Sweden)

    Wen-Chin Chen

    2015-10-01

    Full Text Available As the competitive market nowadays shortens the life cycle of products, new products should be designed to meet the customer’s demand under a dynamic marketing environment so as to efficiently enhance the product strength of new products with maximization of profit. Hence, the key solution for enterprises to succeed will be the precise evaluation of new product development (NPD performance, particularly for those who manage to survive in the intensely competitive market in Taiwan. This study is to identify the thin film transistor-liquid crystal display (TFT-LCD touch panel industry and then establish an integrated model of NPD performance evaluation for enterprises. Firstly, literature review and interviews with experts are conducted to select the four aspects and 15 criteria as the main factors affecting NDP performance evaluation. Secondly, Decision Making Trial and Evaluation Laboratory (DEMATEL is employed to identify the interrelationships among those factors. Finally, a Fuzzy theory is applied to resolve the linguistic hedges and an Analytic Network Process (ANP is adopted to obtain the weights of all factors. A case study is performed to validate the proposed model in a Taiwanese TFT-LCD company. It not only provides the decision maker with a guidance system but also increases the competitive advantages for the TFT-LCD industry to design new products in the future.

  18. EBR-2 [Experimental Breeder Reactor-2] test programs

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  19. BR2 Reactor: Irradiation of fuels

    International Nuclear Information System (INIS)

    Verwimp, A.

    2005-01-01

    Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

  20. Ageing management of the BR2 research reactor

    International Nuclear Information System (INIS)

    Verpoortem, J. R.; Van Dyck, S.

    2014-01-01

    At the Belgian nuclear research centre (SCK.CEN) several test reactors are operated. Among these, Belgian Reactor 2 (BR2) is the largest Material Test Reactor (MTR). This water-cooled, beryllium moderated reactor with a maximum thermal power of 100 MW became operational in 1962. Except for two major refurbishment campaigns of one year each, this reactor has been operated continuously over the past 50 years, with a frequency of 5-12 cycles per year. At present, BR2 is used for different research activities, the production of medical isotopes, the production of n-doped silicon and various training and education activities. (Author)

  1. Trefoil bundles of NPD 7-element size fuel irradiated to 9100 MWd/tonne U

    Energy Technology Data Exchange (ETDEWEB)

    Bain, A S; Christie, J; Daniel, A R

    1964-01-15

    NPD prototype elements (1 in. OD, 19 in. long) were assembled into trefoil bundles and irradiated in the X-5 pressurized-water loop of NRX. The first tests were for only a few weeks but showed that elements made by sheathing UO{sub 2} pellets in Zircaloy-2 behaved well under irradiation; later similar elements were irradiated for 18000 hours to a burn-up of 9100 MWd/tonne U at {integral}kd{theta} = 40 W/cm. The dimensional stability of all the elements was good. Only those subjected to long irradiation showed progressive diametral increases, and these were attributed to relocation of the UO{sub 2} during interim inspections. Length measurements demonstrated that pellet end-dishing is effective in controlling axial expansion, but that for a given depth of dishing the amount of expansion depends on the shoulder width. The extent of grain growth in the UO{sub 2} was compatible with previously reported results when the duration of irradiation, density of the fuel, and variations in growth characteristics of the different batches of UO{sub 2} are considered. The elements taken to high irradiation released up to 135 ml of fission-product gases, which is 2% of the amount formed. The transverse tensile strength of ring samples from the Zircaloy-2 sheaths increased from 75000 to 95000 lb/in{sup 2} at room temperature, but the ductility dropped. The completely brittle fracture of some rings was due to ZrH{sub 2} precipitation. The failure of one element was caused by increased stress due to a higher heat rating, combined with low ductility of the Zircaloy-2 resulting from radiation damage and with precipitation of ZrH{sub 2} because of a lower coolant temperature. The fission-product release from the split was not excessive, and the element was easily withdrawn from the loop after operating at full power for four days from the time of the failure. (author)

  2. 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

  3. BR2 reactor neutron beams

    International Nuclear Information System (INIS)

    Neve de Mevergnies, M.

    1977-01-01

    The use of reactor neutron beams is becoming increasingly more widespread for the study of some properties of condensed matter. It is mainly due to the unique properties of the ''thermal'' neutrons as regards wavelength, energy, magnetic moment and overall favorable ratio of scattering to absorption cross-sections. Besides these fundamental reasons, the impetus for using neutrons is also due to the existence of powerful research reactors (such as BR2) built mainly for nuclear engineering programs, but where a number of intense neutron beams are available at marginal cost. A brief introduction to the production of suitable neutron beams from a reactor is given. (author)

  4. 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.

  5. Medi-Effect Study of NPD Organizational Environment between Transformational Leadership and Green Products Innovation in Green Entrepreneurs%绿色新创企业NPD组织环境对变革型领导与绿色产品创新的中介关系研究

    Institute of Scientific and Technical Information of China (English)

    李先江

    2018-01-01

    When new enterprises carry out green business strategy,in order to explore medi-effect study of NPD organiza-tional environment between transformational leadership and green products innovation this study divide NPD organizational environment into compatibility & importance and resource synergy,build catenation model of transformational leader-ship———organizational environment of new product development———green product innovation,219 firms from four cities of China were studied as empirical samples.The results show that transformational leadership can indirectly controlled by two types of innovation through controlling NPD organization environment.when organizations need to develop explorato-ry green product innovation,transformational leadership can promote the R&D of exploratory green product innovation by strengthening and highlighting the compatibility and importance of NPD organization environments;when organizations need to develop exploitative green product innovation,transformational leadership can promote the R&D of exploitative green product innovation by strengthening and highlighting resource synergy of NPD organization environments.%为了探索实施绿色创业战略时变革型领导在利用式绿色产品创新和探索式绿色产品创新中的作用,将新产品开发(NPD)组织环境作为变革型领导与绿色产品创新间的中介变量,将 NPD 组织环境划分为兼容性与重要性和资源综效,构建了变革型领导 -新产品开发组织环境 -绿色产品创新的连锁关系模型,并以国内4个城市218家企业为样本进行实证分析.研究发现:变革型领导可以通过对新产品开发组织环境的掌控间接控制利用式绿色产品创新和探索式绿色产品创新两类

  6. Pressurized water reactor simulator. Workshop material. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

  7. 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.

  8. Sampling and Analytical Method for Alpha-Dicarbonyl Flavoring Compounds via Derivatization with o-Phenylenediamine and Analysis Using GC-NPD

    Directory of Open Access Journals (Sweden)

    Stephanie M. Pendergrass

    2016-01-01

    Full Text Available A novel methodology is described for the sampling and analysis of diacetyl, 2,3-pentanedione, 2,3-hexanedione, and 2,3-heptanedione. These analytes were collected on o-phenylenediamine-treated silica gel tubes and quantitatively recovered as the corresponding quinoxaline derivatives. After derivatization, the sorbent was desorbed in 3 mL of ethanol solvent and analyzed using gas chromatography/nitrogen-phosphorous detection (GC/NPD. The limits of detection (LOD achieved for each analyte were determined to be in the range of 5–10 nanograms/sample. Evaluation of the on-tube derivatization procedure indicated that it is unaffected by humidities ranging from 20% to 80% and that the derivatization procedure was quantitative for analyte concentrations ranging from 0.1 μg to approximately 500 μg per sample. Storage stability studies indicated that the derivatives were stable for 30 days when stored at both ambient and refrigerated temperatures. Additional studies showed that the quinoxaline derivatives were quantitatively recovered when sampling up to a total volume of 72 L at a sampling rate of 50 cc/min. This method will be important to evaluate and monitor worker exposures in the food and flavoring industry. Samples can be collected over an 8-hour shift with up to 288 L total volume collected regardless of time, sampling rate, and/or the effects of humidity.

  9. Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

    International Nuclear Information System (INIS)

    Bussac, J.; Reuss, P.

    1985-01-01

    This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

  10. TPDWR2: thermal power determination for Westinghouse reactors, Version 2. User's guide

    International Nuclear Information System (INIS)

    Kaczynski, G.M.; Woodruff, R.W.

    1985-12-01

    TPDWR2 is a computer program which was developed to determine the amount of thermal power generated by any Westinghouse nuclear power plant. From system conditions, TPDWR2 calculates enthalpies of water and steam and the power transferred to or from various components in the reactor coolant system and to or from the chemical and volume control system. From these results and assuming that the reactor core is operating at constant power and is at thermal equilibrium, TPDWR2 calculates the thermal power generated by the reactor core. TPDWR2 runs on the IBM PC and XT computers when IBM Personal Computer DOS, Version 2.00 or 2.10, and IBM Personal Computer Basic, Version D2.00 or D2.10, are stored on the same diskette with TPDWR2

  11. Once-through CANDU reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235 U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given

  12. Radiation protection at the RA Reactor in 1988, Part -2, RA reactor annual report

    International Nuclear Information System (INIS)

    Ninkovic, M.; Ajdacic, N.; Zaric, M.; Vukovic, Z.

    1988-01-01

    Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor and radiation protection; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Decontamination and relevant actions, collecting and treatment of fluid effluents; and and solid radioactive wastes [sr

  13. Irradiation effects on Zr-2.5Nb in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

  14. IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

    The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1.

  15. IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    1992-01-01

    The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1

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

    International Nuclear Information System (INIS)

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

    1981-05-01

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

  17. Radiation protection at the RA Reactor in 1998, RA reactor annual report, Part -2

    International Nuclear Information System (INIS)

    Ninkovic, M.; Pavlovic, R.; Mandic, M.; Pavlovic, S.; Grsic, Z.

    1998-01-01

    Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Collecting and treatment of fluid effluents; and (4) radioactive wastes, decontamination and actions. Each of the category is described as a separate annex of this report [sr

  18. Updating and improving the National Population Database to National Population Database 2

    OpenAIRE

    SMITH, Graham; FAIRBURN, Jonathan

    2008-01-01

    In 2004 Staffordshire University delivered the National Population Database for use in estimating populations at risk under the Control of Major Accident Hazards Regulations (COMAH). In 2006 an assessment of the updating and potential improvements to NPD was delivered to HSE. Between Autumn 2007 and Summer 2008 an implementation of the feasibility report led to the creation of National Population Database 2 which both updated and expanded the datasets contained in the original NPD. This repor...

  19. Active species in a large volume N2-O2 post-discharge reactor

    International Nuclear Information System (INIS)

    Kutasi, K; Pintassilgo, C D; Loureiro, J; Coelho, P J

    2007-01-01

    A large volume post-discharge reactor placed downstream from a flowing N 2 -O 2 microwave discharge is modelled using a three-dimensional hydrodynamic model. The density distributions of the most populated active species present in the reactor-O( 3 P), O 2 (a 1 Δ g ), O 2 (b 1 Σ g + ), NO(X 2 Π), NO(A 2 Σ + ), NO(B 2 Π), NO 2 (X), O 3 , O 2 (X 3 Σ g - ) and N( 4 S)-are calculated and the main source and loss processes for each species are identified for two discharge conditions: (i) p = 2 Torr, f = 2450 MHz, and (ii) p = 8 Torr, f = 915 MHz; in the case of a N 2 -2%O 2 mixture composition and gas flow rate of 2 x 10 3 sccm. The modification of the species relative densities by changing the oxygen percentage in the initial gas mixture composition, in the 0.2%-5% range, are presented. The possible tuning of the species concentrations in the reactor by changing the size of the connecting afterglow tube between the active discharge and the large post-discharge reactor is investigated as well

  20. G 2 reactor project; Projet de pile a double fin: G 2

    Energy Technology Data Exchange (ETDEWEB)

    Ailleret, [Electricite de France (EDF), Dir. General des Etudes de Recherches, 75 - Paris (France); Taranger, P; Yvon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    The CEA actually constructs the G-2 reactor core working with natural uranium, which will use graphite as moderator, and gas under pressure as cooling fluid. This report presents the specificity of the new reactor: - the different elements of the reactor core, - the control and the security of the reactor, - the renewal of the fuel, - the biologic surrounding wall, - and the cooling circuit. (M.B.) [French] le Commissariat a l'Energie Atomique construit actuellement la pile G-2 a Uranium naturel, qui utilisera le graphite comme moderateur, et le gaz sous pression comme fluide de refroidissement. Ce rapport presente les specificite du nouveau reacteur: - les differents elements de la pile, - le controle et la securite du reacteur, - le renouvellement du combustible, - l'enceinte biologique, - et le circuit de refroidissement. (M.B.)

  1. Reactor handbook. 2. rev. ed.

    International Nuclear Information System (INIS)

    Lederer, B.J.; Wildberg, D.W.

    1992-01-01

    On the basis of the guidelines on expert knowledge, the book discusses the subjects of atomic physics, heat transfer, nuclear power plants, reactor materials, radiation protection, reactor safety, reactor instrumentation, and reactor operation, with special regard to nuclear power plants with LWR-type reactors. The book is intended for shift personnel, especially gang bosses, reactor operators, and control station operators: for this reason a practical and rather popular style has been chosen. However, the book will also be a manual for other operating personnel, personnel of producer companies, expert organisations, authorities, and students. It can be used as a textbook for staff training, a manual for the practice, and as accompanying book for teaching at nuclear engineering schools. (orig.) With 173 figs [de

  2. Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

    Energy Technology Data Exchange (ETDEWEB)

    Montes Ponce de Leon, J.

    1974-07-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  3. Optimized Control Rods of the BR2 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalcheva, Silva; Koonen, E.

    2007-09-15

    At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

  4. Optimized Control Rods of the BR2 Reactor

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, E.

    2007-01-01

    At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

  5. Upgrading of the research reactors FRG-1 and FRG-2

    International Nuclear Information System (INIS)

    Krull, W.

    1981-01-01

    In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

  6. Annual report on JEN-1 and JEN-2 Reactors

    International Nuclear Information System (INIS)

    Montes Ponce de Leon, J.

    1974-01-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  7. System Design of a Supercritical CO_2 cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Kim, Seong Gu; Cho, Seongkuk; Yu, Hwanyeal; Kim, Yonghee; Jeong, Yong Hoon; Lee, Jeong Ik

    2014-01-01

    Small modular reactor (SMR) systems that have advantages of little initial capital cost and small restriction on construction site are being developed by many research organizations around the world. Existing SMR concepts have the same objective: to achieve compact size and a long life core. Most of small modular reactors have much smaller size than the large nuclear power plant. However, existing SMR concepts are not fully modularized. This paper suggests a complete modular reactor with an innovative concept for reactor cooling by using a supercritical carbon dioxide. The authors propose the supercritical CO_2 Brayton cycle (S-CO_2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the reactor core and PCU in one vessel. A conceptual design of the proposed small modular reactor was developed, which is named as KAIST Micro Modular Reactor (MMR). The supercritical CO_2 Brayton cycle for the S-CO_2 cooled reactor core was optimized and the size of turbomachinery and heat exchanger were estimated preliminary. The nuclear fuel composed with UN was proposed and the core lifetime was obtained from a burnup versus reactivity calculation. Furthermore, a system layout with fully passive safety systems for both normal operation and emergency operation was proposed. (author)

  8. The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

    International Nuclear Information System (INIS)

    Townes, B.M.; Hilborn, J.W.

    1985-06-01

    A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

  9. Operation of the SLOWPOKE-2 reactor in Jamaica

    Energy Technology Data Exchange (ETDEWEB)

    Grant, C.N.; Lalor, G.C.; Vuchkov, M.K. [University of the West Indies, Kingston (Jamaica)

    2001-07-01

    Over the past sixteen years lCENS has operated a SLOWPOKE 2 nuclear reactor almost exclusively for the purpose of neutron activation analysis. During this period we have adopted a strategy of minimum irradiation times while optimizing our output in an effort to increase the lifetime of the reactor core and to maintaining fuel integrity. An inter-comparison study with results obtained with a much larger reactor at IPEN has validated this approach. The parameters routinely monitored at ICENS are also discussed and the method used to predict the next shim adjustment. (author)

  10. Irradiation techniques at BR2 reactor

    International Nuclear Information System (INIS)

    Hebel, W.

    1978-01-01

    Since 1963 the material testing reactor BR2 at Mol is operated for the realisation of numerous research programs and experiments on the behavior of materials under nuclear radiation and in particular under intensive neutron exposure. During this period special irradiation techniques and experimental devices were developed according to the desiderata of the different experiments and to the irradiation possibilities offered at BR2. The design and the operating characteristics of quite a number of those irradiation rigs of proven reliability may be used or can be made available for new irradiation experiments. A brief description is given of some typical irradiation devices designed and constructed by CEN/SCK, Technology and Energy Dpt. They are compiled according to their main use for the different research and development programs realized at BR2. Their eventual application however for different objectives could be possible. A final chapter summarizes the principal irradiation conditions offered by BR2 reactor. (author)

  11. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1985-09-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  12. TMI-2 reactor vessel plenum final lift

    International Nuclear Information System (INIS)

    Wilson, D.C.

    1986-01-01

    Removal of the plenum assembly from the TMI-2 reactor vessel was necessary to gain access to the core region for defueling. The plenum was lifted from the reactor vessel by the polar crane using three specially designed pendant assemblies. It was then transferred in air to the flooded deep end of the refueling canal and lowered onto a storage stand where it will remain throughout the defueling effort. The lift and transfer were successfully accomplished on May 15, 1985 in just under three hours by a lift team located in a shielded area within the reactor building. The success of the program is attributed to extensive mockup and training activities plus thorough preparations to address potential problems. 54 refs

  13. Refurbishment programme for the BR2-reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koonen, E [Centre d' Etude de l' Energie Nucleaire, Studiecentrum voor Kernenergie, BR2 Department, Boeretang, Mol (Belgium)

    1992-07-01

    BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

  14. Refurbishment programme for the BR2-reactor

    International Nuclear Information System (INIS)

    Koonen, E.

    1992-01-01

    BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

  15. Refurbishing the BR2 materials testing reactor

    International Nuclear Information System (INIS)

    Baugnet, J.M.; Dekeyser, J.; Gubel, P.

    1995-01-01

    SCK/CEN is refurbishing its BR2 reactor to allow its further operation during the next 15 years; in doing so, it chooses to keep BR2 available for future scientific and technological irradiation programs within an international context. (author) 2 figs

  16. Four Novel p.N385K, p.V36A, c.1033–1034insT and c.1417–1418delCT Mutations in the Sphingomyelin Phosphodiesterase 1 (SMPD1 Gene in Patients with Types A and B Niemann-Pick Disease (NPD

    Directory of Open Access Journals (Sweden)

    Masoumeh Dehghan Manshadi

    2015-03-01

    Full Text Available Background: Types A and B Niemann-Pick disease (NPD are autosomal-recessive lysosomal storage disorders caused by the deficient activity of acid sphingomyelinase due to mutations in the sphingomyelin phosphodiesterase 1 (SMPD1 gene. Methods: In order to determine the prevalence and distribution of SMPD1 gene mutations, the genomic DNA of 15 unrelated Iranian patients with types A and B NPD was examined using PCR, DNA sequencing and bioinformatics analysis. Results: Of 8 patients with the p.G508R mutation, 5 patients were homozygous, while the other 3 were heterozygous. One patient was heterozygous for both the p.N385K and p.G508R mutations. Another patient was heterozygous for both the p.A487V and p.G508R mutations. Two patients (one homozygous and one heterozygous showed the p.V36A mutation. One patient was homozygous for the c.1033–1034insT mutation. One patient was homozygous for the c.573delT mutation, and 1 patient was homozygous for the c.1417–1418delCT mutation. Additionally, bioinformatics analysis indicated that two new p.V36A and p.N385K mutations decreased the acid sphingomyelinase (ASM protein stability, which might be evidence to suggest the pathogenicity of these mutations. Conclusion: with detection of these new mutations, the genotypic spectrum of types A and B NPD is extended, facilitating the definition of disease-related mutations. However, more research is essential to confirm the pathogenic effect of these mutations.

  17. Proceedings of 2. Yugoslav symposium on reactor physics, Part 2, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 2 of the Proceedings of 2. Yugoslav symposium on reactor physics includes eight papers dealing with the following topics: method for measuring high anti reactivities of a reactor system; integration method for thermal reaction rate calculation; Determination of initial core configuration for BHWR-200 MWe; safety shutdowns and failures of the RA reactor equipment; determining the reactivity of absorption rods; measurements of thermal and fast neutron fluxes at the TRIGA reactor and other measurements during operation of the TRIGA reactor; mathematical modelling of the reactor safety; review of problems and methods for radiation risk assessment in the environment of a nuclear power plant

  18. SIRIUS 2: A versatile medium power research reactor

    International Nuclear Information System (INIS)

    Rousselle, P.

    1992-01-01

    Most of the Research Reactors in the world have been critical in the Sixties and operated for twenty to thirty years. Some of them have been completely shut down, modified, or simply refurbished; the total number of RR in operation has decreased but there is still an important need for medium power research reactors in order: - to sustain a power program with fuel and material testing for NPP or fusion reactors; - to produce radioisotopes for industrial or medical purposes, doped silicon, NAA or neutron radiography; - to investigate further the condensed matter, with cold neutrons routed through neutron guides to improved equipment; - to develop new technologies and applications such as medical alphatherapy. Hence, taking advantage of nearly hundred reactor x years operation and backed up by the CEA experience, TECHNICATOME assisted by FRAMATOME has designed a new versatile multipurpose Research Reactor (20-30 Mw) SIRIUS 2 taking into account: - more stringent safety rules; - the lifetime; - the flexibility enabling a wide range of experiments and, - the future dismantling of the facility according to the ALARA criteria

  19. OTUS - Reactor inventory management system based on ORIGEN2

    Energy Technology Data Exchange (ETDEWEB)

    Poellaenen, R; Toivonen, H; Lahtinen, J; Ilander, T

    1995-10-01

    ORIGEN2 is a computer code that calculates nuclide composition and other characteristics of nuclear fuel. The use of ORIGEN2 requires good knowledge in reactor physics. However, once the input has been defined for a particular reactor type, the calculations can be easily repeated for any burnup and decay time. This procedure produces large output files that are difficult to handle manually. A new computer code, known as OTUS, was designed to facilitate the postprocessing of the data. OTUS makes use of the inventory files precalculated with ORIGEN2 in a way that enables their versatile treatment for different safety analysis purposes. A data base is created containing a comprehensive set of ORIGEN2 calculations as a function of fuel burnup and decay time. OTUS is a reactor inventory management system for a microcomputer with Windows interface. Four major data operations are available: (1) Build data modifies ORIGEN2 output data into a suitable format, (2) View data enables flexible presentation of the data as such, (3) Different calculations, such as nuclide ratios and hot particle characteristics, can be performed for severe accident analyses, consequence analyses and research purposes, (4) Summary files contain both burnup dependent and decay time dependent inventory information related to the nuclide and the reactor specified. These files can be used for safeguards, radiation monitoring and safety assessment. (orig.) (22 refs., 29 figs.).

  20. The 2nd reactor core of the NS Otto Hahn

    International Nuclear Information System (INIS)

    Manthey, H.J.; Kracht, H.

    1979-01-01

    Details of the design of the 2nd reactor core are given, followed by a brief report summarising the operating experience gained with this 2nd core, as well as by an evaluation of measured data and statements concerning the usefulness of the knowledge gained for the development of future reactor cores. Quite a number of these data have been used to improve the concept and thus the specifications for the fuel elements of the 3rd core of the reactor of the NS Otto Hahn. (orig./HP) [de

  1. EL-2 reactor: Thermal neutron flux distribution

    International Nuclear Information System (INIS)

    Rousseau, A.; Genthon, J.P.

    1958-01-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  2. Physics design of advanced steady-state tokamak reactor A-SSTR2

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Ushigusa, Kenkichi

    2000-10-01

    Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (R p =6.2m, a p =1.5m, I p =12MA) has a fusion power of 4GW with β N =4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)

  3. FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients

    International Nuclear Information System (INIS)

    Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.

    1984-01-01

    1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry

  4. Keeping research reactors relevant: A pro-active approach for SLOWPOKE-2

    International Nuclear Information System (INIS)

    Cosby, L.R.; Bennett, L.G.I.; Nielsen, K.; Weir, R.

    2010-01-01

    The SLOWPOKE is a small, inherently safe, pool-type research reactor that was engineered and marketed by Atomic Energy of Canada Limited (AECL) in the 1970s and 80s. The original reactor, SLOWPOKE-1, was moved from Chalk River to the University of Toronto in 1970 and was operated until upgraded to the SLOWPOKE-2 reactor in 1973. In all, eight reactors in the two versions were produced and five are still in operation today, three having been decommissioned. All of the remaining reactors are designated as SLOWPOKE-2 reactors. These research reactors are prone to two major issues: aging components and lack of relevance to a younger audience. In order to combat these problems, one SLOWPOKE -2 facility has embraced a strategy that involves modernizing their reactor in order to keep the reactor up to date and relevant. In 2001, this facility replaced its aging analogue reactor control system with a digital control system. The system was successfully commissioned and has provided a renewed platform for student learning and research. The digital control system provides a better interface and allows flexibility in data storage and retrieval that was never possible with the analogue control system. This facility has started work on another upgrade to the digital control and instrumentation system that will be installed in 2010. The upgrade includes new computer hardware, updated software and a web-based simulation and training system that will allow licensed operators, students and researchers to use an online simulation tool for training, education and research. The tool consists of: 1) A dynamic simulation for reactor kinetics (e.g., core flux, power, core temperatures, etc). This tool is useful for operator training and student education; 2) Dynamic mapping of the reactor and pool container gamma and neutron fluxes as well as the vertical neutron beam tube flux. This research planning tool is used for various researchers who wish to do irradiations (e.g., neutron

  5. Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

    International Nuclear Information System (INIS)

    Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

    1983-02-01

    The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems

  6. Thermal reactor benchmark tests on JENDL-2

    International Nuclear Information System (INIS)

    Takano, Hideki; Tsuchihashi, Keichiro; Yamane, Tsuyoshi; Akino, Fujiyoshi; Ishiguro, Yukio; Ido, Masaru.

    1983-11-01

    A group constant library for the thermal reactor standard nuclear design code system SRAC was produced by using the evaluated nuclear data JENDL-2. Furthermore, the group constants for 235 U were calculated also from ENDF/B-V. Thermal reactor benchmark calculations were performed using the produced group constant library. The selected benchmark cores are two water-moderated lattices (TRX-1 and 2), two heavy water-moderated cores (DCA and ETA-1), two graphite-moderated cores (SHE-8 and 13) and eight critical experiments for critical safety. The effective multiplication factors and lattice cell parameters were calculated and compared with the experimental values. The results are summarized as follows. (1) Effective multiplication factors: The results by JENDL-2 are considerably improved in comparison with ones by ENDF/B-IV. The best agreement is obtained by using JENDL-2 and ENDF/B-V (only 235 U) data. (2) Lattice cell parameters: For the rho 28 (the ratio of epithermal to thermal 238 U captures) and C* (the ratio of 238 U captures to 235 U fissions), the values calculated by JENDL-2 are in good agreement with the experimental values. The rho 28 (the ratio of 238 U to 235 U fissions) are overestimated as found also for the fast reactor benchmarks. The rho 02 (the ratio of epithermal to thermal 232 Th captures) calculated by JENDL-2 or ENDF/B-IV are considerably underestimated. The functions of the SRAC system have been continued to be extended according to the needs of its users. A brief description will be given, in Appendix B, to the extended parts of the SRAC system together with the input specification. (author)

  7. WWER-1000 reactor simulator. Material for training courses and workshops. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No.12, Reactor Simulator Development (2001). Course material for workshops using a pressurized water reactor (PWR) simulator developed for the IAEA by Cassiopeia Technologies Inc. of Canada is presented in the IAEA publication, Training Course Series No. 22, 2nd edition, Pressurized Water Reactor Simulator (2005) and Training Course Series No.23, 2nd edition, Boiling Water Reactor Simulator (2005). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation

  8. ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

    International Nuclear Information System (INIS)

    Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

    2010-01-01

    The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

  9. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 2-2

    International Nuclear Information System (INIS)

    Okuda, Eiji; Ito, Hiromichi; Yoshihara, Shizuya

    2014-01-01

    An accident occurred in experimental fast reactor 'Joyo' in 2007 which is obstruction of fuel change equipment caused by contacting rotating plug and MARICO-2. In addition, we confirmed two happenings in the reactor vessel that (1) Deformation of MARICO-2 subassembly on the in vessel storage rack together with a transfer pot, (2) Deformation of the Upper core structure of 'Joyo' caused by contacting MARICO-2 subassembly and the UCS. We do the restoration work for restoring it. This time, we describe current status of Replacement work of the UCS. (author)

  10. Sterilization of E. coli bacterium in a flowing N2-O2 post-discharge reactor

    International Nuclear Information System (INIS)

    Villeger, S; Cousty, S; Ricard, A; Sixou, M

    2003-01-01

    Effective destruction of Escherichia coli (E. coli) bacteria has been obtained in a flowing N 2 -O 2 microwave post-discharge reactor. The sterilizing agents are the O atoms and the UV emissions of NOβ which are produced by N and O atoms recombination in the reactor. In the following plasma conditions: pressure 5 Torr, flow rate 1 L n min -1 , microwave power of 100 W in a quartz tube of 5 mm, an O atom density of 2.5x10 15 cm -3 is measured by NO titration in the post-discharge reactor with UV emission in a N 2 -(5%)O 2 gas mixture. Full destruction of 10 13 cfu ml -1 E. coli is observed after a treatment time of 25 min. (rapid communication)

  11. Safe dismantling of the SVAFO research reactors R2 and R2-0 in Sweden

    International Nuclear Information System (INIS)

    ARNOLD, Hans-Uwe; BROY, Yvonne; Dirk Schneider

    2017-01-01

    The R2 and R2-0 reactors were part of the Swedish government's research program on nuclear power from the early 1960's. Both reactors were shut down in 2005 following a decision by former operator Studsvik Nuclear AB. The decommissioning of the R2 and R2-0 reactors is divided into three phases. The first phase - awarded to AREVA - involved dismantling of the reactors and associated systems in the reactor pool, treatment of the disassembled components as well as draining, cleaning and emptying the pool. In the second phase, the pool structure itself will be dismantled, while removal of remaining reactor systems, treatment and disposal of materials and clean-up will be carried out in the third stage. The entire work is planned to be completed before the end of this decade. The paper describes the several steps of phase 1 - starting with the team building, followed by the dismantling operations and covers challenges encountered and lessons learned as well. The reactors consist of 5.400 kg aluminum, 6.000 kg stainless steel restraint structures as well as, connection elements of the mostly flanged components (1.000 kg). The most demanding - from a radiological point of view - was the R2-0 reactor that was limited to ∼ 1 m"3 construction volumes but with an extremely heterogeneous activation profile. Based on the calculated radiological entrance data and later sampling, nuclide vectors for both reactors depending on the real placement of the single component and on the material (aluminum and stainless steel) were created. Finally, for the highest activated component from R2 reactor, 85 Sv/h were measured. The dismantling principles - adopted on a safety point of view - were the following: The always protected base area of the ponds served as a flexible buffer area for waste components and packaging. Specific protections were also installed on the walls to protect them from mechanical stress which may occur during dismantling work. A specific work platform was

  12. Benchmark calculations for VENUS-2 MOX -fueled reactor dosimetry

    International Nuclear Information System (INIS)

    Kim, Jong Kung; Kim, Hong Chul; Shin, Chang Ho; Han, Chi Young; Na, Byung Chan

    2004-01-01

    As a part of a Nuclear Energy Agency (NEA) Project, it was pursued the benchmark for dosimetry calculation of the VENUS-2 MOX-fueled reactor. In this benchmark, the goal is to test the current state-of-the-art computational methods of calculating neutron flux to reactor components against the measured data of the VENUS-2 MOX-fuelled critical experiments. The measured data to be used for this benchmark are the equivalent fission fluxes which are the reaction rates divided by the U 235 fission spectrum averaged cross-section of the corresponding dosimeter. The present benchmark is, therefore, defined to calculate reaction rates and corresponding equivalent fission fluxes measured on the core-mid plane at specific positions outside the core of the VENUS-2 MOX-fuelled reactor. This is a follow-up exercise to the previously completed UO 2 -fuelled VENUS-1 two-dimensional and VENUS-3 three-dimensional exercises. The use of MOX fuel in LWRs presents different neutron characteristics and this is the main interest of the current benchmark compared to the previous ones

  13. PCU arrangement of a supercritical CO{sub 2} cooled micro modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong Gu; Baik, Seungjoon; Cho, Seong Kuk; Oh, Bong Seong; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    As part of the SMR(Small Modular Reactor)s development effort, the authors propose a concept of supercritical CO{sub 2} (S-CO{sub 2}) cooled fast reactor combined with the S-CO{sub 2} Brayton cycle. The reactor concept is named as KAIST Micro Modular Reactor (MMR). The S-CO{sub 2} Brayton cycle has many strong points when it is used for SMR's power conversion unit. It occupies small footprints due to the compact cycle components and simple layout. Thus, a concept of one module containing the S-CO{sub 2} cooled fast reactor and power conversion system is possible. This module can be shipped via ground transportation (by trailer) or marine transportation. In this study, the authors propose a new conceptual layout for the S-CO{sub 2} cooled direct cycle while considering various issues for arranging cycle components. The new design has an improved cycle efficiency (from 31% to 34%) than the earlier version of MMR by reducing pressure drops in the heat exchangers. As a more efficient option, a recompression recuperated cycle was also designed. It improves 5% of thermal efficiency while 18tons of mass can be added in comparison to the simple recuperated cycle. Even if we adopt recompression cycle as a PCU, the weight of module (152tons) is less than the ground transportable limit (260tons)

  14. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1984-12-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  15. Neutronic study using oxide and nitride fuels for the Super Phenix 2 reactor

    International Nuclear Information System (INIS)

    Batista, J.L.; Renke, C.A.C.

    1991-11-01

    This report presents a neutronic analysis and a description of the Super Phenix 2 reactor, taken as reference. We present the methodology and results for cell and global reactor calculations for oxide (U O 2 - Pu O 2 ) and nitride (U N - Pu N) fuels. To conclude we compare the performance of oxide and nitride fuels for the reference reactor. (author)

  16. Shadow corrosion evaluation in the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Sanders, Ch.; Lysell, G.

    2000-01-01

    Post-irradiation examination has shown that increased corrosion occurs when zirconium alloys are in contact with or in proximity to other metallic objects. The observations indicate an influence of irradiation from the adjacent component as the enhanced corrosion occurs as a 'shadow' of the metallic object on the zirconium surface. This phenomenon could ultimately limit the lifetime of certain zirconium alloy components in the reactor. The Studsvik R2 materials test reactor has an In-Core Autoclave (INCA) test facility especially designed for water chemistry and materials research. The INCA facility has been evaluated and found suitable for shadow corrosion studies. The R2 reactor core containing the INCA facility was modeled with the Monte Carlo N-Particle (MCNP) code in order to evaluate the electron deposition in various materials and to develop a hypothesis of the shadow corrosion mechanism. (authors)

  17. Engineering strain measurements using the NPD at LANSCE

    International Nuclear Information System (INIS)

    Bourke, M.A.M.; Goldstone, J.A.; Lovell, K.J.

    1991-01-01

    The presence of residual stress in engineering components can affect their mechanical properties and structural integrity. Neutron diffraction is the only measuring technique which can provide spatially resolved non-destructive strain measurements in the interior of a component. By recording the change in the interplanar spacings elastic strains can be measured for individual lattice reflections. Also on a pulsed source, where all lattice reflections are recorded, profile refinement is an option which allows the strain to be obtained from changes in the lattice parameter. Measurements made at LANSCE demonstrate the potential for stress measurements on a pulsed source and indicate the advantages and disadvantages over measurements made on a reactor. (author)

  18. Benchmark tests of JENDL-3.2 for thermal and fast reactors

    International Nuclear Information System (INIS)

    Takano, Hideki

    1995-01-01

    Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

  19. Radionuclide distribution in TMI-2 reactor building basement liquids and solids

    International Nuclear Information System (INIS)

    Horan, J.T.; McIsaac, C.V.; Keefer, D.G.

    1984-01-01

    As a result of the TMI-2 accident, approximately 2.46 x 10 6 L of contaminated water were released to the Reactor Building basement. The principal fission product release pathway from the damaged core was through the reactor coolant system (RCS) to the pressurizer, through the pressure-operated relief valve (PORV) on the pressurizer to the Reactor Coolant Drain Tank (RCDT), and then through the RCDT rupture disk to the Reactor Building basement. Since August 1979, a number of efforts have been made to determine the location, quantity, and composition of fission products released to the Reactor Building basement. These efforts have included sampling of the basement water and solids, the basement sump pump recirculation line, the RCDT, and visual surveys using a closed circuit television (CCTV) system. The analysis of basement samples has provided data on the physical and radioisotopic characteristics of the liquids and solids. This paper describes the sample collection techniques and discusses radiochemical analyses results

  20. An experimental investigation of fission product release in SLOWPOKE-2 reactors

    International Nuclear Information System (INIS)

    Harnden, A.M.C.

    1995-09-01

    Increasing radiation fields due to a release of fission products in the reactor container of several SLOWPOKE-2 reactors fuelled with a highly-enriched uranium (HEU) alloy core have been observed. It is believed that these increases are associated with the fuel fabrication where a small amount of uranium-bearing material is exposed to the coolant at the end-welds of the fuel element. To investigate this phenomenon samples of reactor water and gas from the headspace above the water have been obtained and examined by gamma spectrometry methods for reactors of various burnups at the University of Toronto, Ecole Polytechnique and Kanata Isotope Production Facility. An underwater visual examination of the fuel core at Ecole Polytechnique has also provided information on the condition of the core. This report (Volume 1) summarizes the equipment, analysis techniques and results of tests conducted at the various reactor sites. The data report is published as Volume 2. (author). 30 refs., 9 tabs., 20 figs

  1. The Oak Ridge Research Reactor: safety analysis: Volume 2, supplement 2

    International Nuclear Information System (INIS)

    Hurt, S.S.

    1986-11-01

    The Oak Ridge Research Reactor Safety Analysis was last updated via ORNL-4169, Vol. 2, Supplement 1, in May of 1978. Since that date, several changes have been effected through the change-memo system described below. While these changes have involved the cooling system, the electrical system, and the reactor instrumentation and controls, they have not, for the most part, presented new or unreviewed safety questions. However, some of the changes have been based on questions or recommendations stemming from safety reviews or from reactor events at other sites. This paper discusses those changes which were judged to be safety related and which include revisions to the syphon-break system and changes related to seismic considerations which were very recently completed. The maximum hypothetical accident postulated in the original safety analysis requires dynamic containment and filtered flow for compliance with 10CFR100 limits at the site boundary

  2. Loss of coolant analysis for the tower shielding reactor 2

    International Nuclear Information System (INIS)

    Radcliff, T.D.; Williams, P.T.

    1990-06-01

    The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

  3. Dalhousie SLOWPOKE-2 reactor: A nuclear analytical chemistry facility

    International Nuclear Information System (INIS)

    Chatt, A.; Holzbecher, J.

    1990-01-01

    SLOWPOKE is an acronym for Safe Low POwer Kritical Experiment. The SOWPOKE-2 is a compact, inherently safe, swimming-pool-type reactor designed by the Atomic Energy of Canada Limited for neutron activation analysis (NAA) and isotope production. The Dalhousie University SLOWPOKE-2 reactor (DUSR) has been operating since 1976; a large beryllium reflector was added in 1986 to extend its lifetime by another 8 to 10 yr. The DUSR is generally operated at half-power with a maximum thermal flux of 1.1 x 10 12 n/cm 2 ·s in the inner pneumatic sites and that of 5.4 x 10 11 n/cm 2 ·s in the outer sites. Despite this comparatively low flux, SLOWPOKE-2 reactors have many beneficial features that are continuously being exploited at the DUSR facility for developing nuclear analytical methods for fundamental as well as applied studies. Although NAA is a well-established analytical technique, much of the activation analysis being performed in most facilities has been limited to methods using fairly long-lived nuclides. The approach at the DUSR facility has been to utilize the highly homogeneous, stable, and reproducible neutron flux to develop NAA methods based on short-lived nuclides. SLOWPOKE reactors have a fairly high epithermal neutron flux, which is being advantageously used for determining several trace elements in complex matrices. Radiochemical NAA (RNAA) methods using coprecipitation, distillation, and ion-exchange separations have been used for the determination of very low levels of several elements in biological materials

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

    International Nuclear Information System (INIS)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B.

    1996-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-09-01

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

  6. Dynamic simulation of the 2 MWt slowpoke heating reactor

    International Nuclear Information System (INIS)

    Tseng, C.M.; Lepp, R.M.

    1982-04-01

    A 2 MWt SLOWPOKE reactor, intended for commercial space heating, is being developed at the Chalk River Nuclear Laboratories. A small-signal dynamic simulation of this reactor, without closed-loop control, was developed. Basic equations were used to describe the physical phenomena in each kf the eight reactor subsystems. These equations were then linearized about the normal operation conditions and rearranged in a dimensionless form for implementation. The overall simulation is non-linear. Slow transient responses (minutes to days) of the simulation to both reactivity and temperature perturbations were measured at full power. In all cases the system reached a new steady state in times varying from 12 h to 250 h. These results illustrate the benefits of the inherent negative reactivity feedback of this reactor concept. The addition of closed-loop control using core outlet temperature as the controlled variable to move a beryllium reflector is also examined

  7. Distribution of energy of impulses of the modernized IBR-2 REACTOR

    International Nuclear Information System (INIS)

    Tayibov, L.A; Mehtiyeva, R.N.; )

    2011-01-01

    Full text: For the modernized IBR-2 reactor there are two main reasons causing fluctuations of energy of impulses [1,3] on low power of stochastic fluctuations, on the nominal - giving rise to fluctuations of external reactance. The fluctuations of pulse energy is quite significant (20%). They affect the dynamics of the reactor, the process of regulation, starting, as well as the work of the experimental apparatus, etc. It is clear that research of fluctuation of energy of impulses has special value for the IBR-2 type reactor. Sufficient information about the statistical properties of the reactor noise gives the density distribution of the energy pulse power. We used the usual procedure of statistical analysis of time series. Calculated pulse energy of density and the parameters of this distribution.

  8. A Complete MCDM Model for NPD Performance Assessment in an LED-Based Lighting Plant Factory

    Directory of Open Access Journals (Sweden)

    Wen-Chin Chen

    2018-01-01

    Full Text Available Globally, industries and economies have undergone rapid development and expansion over the last several decades. As a result, global warming and environmental contaminations have resulted in climate change and jeopardized food security. In many developing countries, already decreasing crop yields are threatened by extreme weather and soil damaged by genetically modified food, making environmental problems worse and increasing food and organic product prices. For these reasons, this study proposes a hybrid multicriteria decision-making (MCDM model for new product development (NPD in the light-emitting diode- (LED- based lighting plant factory. First, literature reviews and expert interviews are employed in constructing a list of decision-making objectives and criteria for new product development. Then, a fuzzy Delphi method (FDM is used to screen the elements of the objectives and criteria, while a fuzzy decision-making trial and evaluation laboratory (FDEMATEL is used to determine the relationships among the objectives and criteria. Finally, a fuzzy analytic network process (FANP and a composite priority vector (CPV are manipulated to determine the relative importance weights of the critical objectives and criteria. Results show that the proposed method can create a useful and assessable MCDM model for decision-making applications in new product development, and a case study is herein performed to validate the feasibility of the proposed model in a Taiwanese LED-based lighting plant factory, which not only provides the decision-makers with a feasible hierarchical data structure for decision-making guidance but also increases the competitive advantages of trade-offs on developing novel products.

  9. Engineering strain measurements using the NPD at LANSCE

    International Nuclear Information System (INIS)

    Bourke, M.A.M.; Goldstone, J.A.; Lovell, K.J.

    1990-01-01

    The presence of residual stress in engineering components can affect their mechanical properties and structural integrity. Neutron diffraction is the only measuring technique which can provide spatially resolved non-destructive strain measurements in the interior of a component. By recording the change in the interplanar spacings elastic strains can be measured for individual lattice reflections. Also on a pulsed source, where all lattice reflections are recorded, profile refinement is an option which alloys the strain to be obtained from changes in the lattice parameter. Measurements made at LANSCE demonstrate the potential for stress measurements on a pulsed source and indicate the advantages and disadvantages over measurements made on a reactor. 5 refs., 5 figs

  10. Equipment for thermal neutron flux measurements in reactor R2

    Energy Technology Data Exchange (ETDEWEB)

    Johansson, E; Nilsson, T; Claeson, S

    1960-04-15

    For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

  11. A conceptual design of LIB fusion reactor: UTLIF(2)

    International Nuclear Information System (INIS)

    Madarame, Haruki; Kondo, Shunsuke; Iwata, Shuichi; Oka, Yoshiaki; Miya, Kenzo.

    1984-01-01

    UTLIF(2) is a conceptual design study on a light ion beam driven fusion reactor based on a concept of rod-bundle blanket. Survivability and maintainability of the first wall and the blanket are regarded as of major importance in the design. The blanket rod is composed of a thick tube which has enough stiffness, a thin wrapping wall which receives high heat flux, and liquid lithium which breeds tritium and removes generated heat. The rod can be pulled out from the outside of the reactor vessel, hence the replacement is very easy. Nuclear and thermal analysis have been made and the performance of the reactor has been shown to be satisfactory. (author)

  12. The BR2 materials testing reactor. Past, ongoing and under-study upgradings

    Energy Technology Data Exchange (ETDEWEB)

    Baugnet, J M; Roedt, Ch de; Gubel, P; Koonen, E [Centre d' Etude de I' Energie Nucleaire, Studiecentrum voor Kernenergie, C.E.N./S.C.K., Mol (Belgium)

    1990-05-01

    The BR2 reactor (Mol, Belgium) is a high-flux materials testing reactor. The fuel is 93% {sup 235}U enriched uranium. The nominal power ranges from 60 to 100 MW. The main features of the design are the following: 1) maximum neutron flux, thermal: 1.2 x 10{sup 15} n/cm{sup 2} s; fast (E > 0.1 MeV) : 8.4 x 10{sup 14} n /cm{sup 2} s; 2) great flexibility of utilization: the core configuration and operation mode can be adapted to the experimental loading; 3) neutron spectrum tailoring; 4) availability of five 200 mm diameter channels besides the standard channels (84 mm diameter); 5) access to the top and bottom covers of the reactor authorizing the irradiation of loops. The reactor is used to study the behaviour of fuel elements and structural materials intended for future nuclear power stations of several types (fission and fusion). Irradiations are carried out in connection with performance tests up to very high burn-up or neutron fluence as well as for safety experiments, power cycling experiments, and generally speaking, tests under off-normal conditions. Irradiations for nuclear transmutation (production of high specific activity radio-isotopes and transplutonium elements), neutron-radiography, use of beam tubes for physics studies, and gamma irradiations are also carried out. The BR2 is used in support of Belgian programs, at the request of utilities, industry and universities and in the framework of international agreements. The paper reviews the past and ongoing upgrading and enhancement of reactor capabilities as well as those under study or consideration, namely with regard to: reactor equipment, fuel elements, irradiation facilities, reactor operation conditions and long-term strategy. (author)

  13. TiO2 Solar Photocatalytic Reactor Systems: Selection of Reactor Design for Scale-up and Commercialization—Analytical Review

    Directory of Open Access Journals (Sweden)

    Yasmine Abdel-Maksoud

    2016-09-01

    Full Text Available For the last four decades, viability of photocatalytic degradation of organic compounds in water streams has been demonstrated. Different configurations for solar TiO2 photocatalytic reactors have been used, however pilot and demonstration plants are still countable. Degradation efficiency reported as a function of treatment time does not answer the question: which of these reactor configurations is the most suitable for photocatalytic process and optimum for scale-up and commercialization? Degradation efficiency expressed as a function of the reactor throughput and ease of catalyst removal from treated effluent are used for comparing performance of different reactor configurations to select the optimum for scale-up. Comparison included parabolic trough, flat plate, double skin sheet, shallow ponds, shallow tanks, thin-film fixed-bed, thin film cascade, step, compound parabolic concentrators, fountain, slurry bubble column, pebble bed and packed bed reactors. Degradation efficiency as a function of system throughput is a powerful indicator for comparing the performance of photocatalytic reactors of different types and geometries, at different development scales. Shallow ponds, shallow tanks and fountain reactors have the potential of meeting all the process requirements and a relatively high throughput are suitable for developing into continuous industrial-scale treatment units given that an efficient immobilized or supported photocatalyst is used.

  14. Planned Scientific programs around the Triga Mark 2 Reactor

    International Nuclear Information System (INIS)

    Majah, M Ibn.

    2007-01-01

    Full text: Nuclear techniques have been introduced to Morocco since the sixties. After the energy crisis of 1973, Morocco decides to create the National Center for Energy Sciences and Nuclear Techniques (CNESTEN) under the supervision of the Ministry of high Education and Research, with a research commercial and support vocation. CNESTEN is in charge of promoting nuclear application, to act as technical support for the authorities and to prepare the technological basis for nuclear power option. In 1998, CNESTEN started the construction of Nuclear Research Centre. The on going activities cover many sectors : earth and environmental sciences, high energy physics, safety and security, waste management. In 2001, CNESTEN started the construction of a 2MW TRiga Mark 2 Reactor, with the possibility to increase the power to 3 MW. The construction was achieved in January 2007. The operation of the reactor is expected for April 2007. The program of the utilization of the reactor was established with th contribution of the university and with the assistance of IAEA. Some of the experimental set-up installed around the reactor have been designed. CNESTEN has developed cooperation with Nuclear research centres from other countries and is receiving visitors and trainees mainly through the IAEA [fr

  15. Update neutron nuclear data evaluation for 236,238Np

    International Nuclear Information System (INIS)

    Chen, Guochang; Wang, Jimin; Yu, Baosheng; Cao, Wentian; Tang, Guo-you

    2015-01-01

    The nuclear data with high accuracy for actinides play an important role in nuclear technology applications, including reactor design and operation, fuel cycle, estimation of the amount of minor actinides (MAs) in high burnup reactors and to research to transmute the MAs to short half-lived nuclides or stable ones. The nuclides of 236 Np are generated via the α-decay of 240 Am or 237 Np(n, 2n) and 237 Np(d, t) reactions. And the nuclides of 238 Np are generated via the α-decay of 242 Am or 237 Np(n, γ) and 237 Np(d, p) reactions. In the present work, according to the systematic trend of the total cross section and elastic cross section etc. of different Np isotopes, and based on the neutron optical model parameters (OMP) of 237 Np, a new set of neutron optical model parameters were obtained for 236,238 Np. Based on the new set OMP and the systematic trend of the cross sections of different Np isotopes, a full set of 236,238 Np neutron nuclear data has been updated and improved by theoretical calculation. The present result has significant improvements over the data in CENDL-3.1

  16. Power noise spectrum classification in the problem of the IBR-2 reactor

    International Nuclear Information System (INIS)

    Bargel, M.; Kitowski, J.; Pepelyshev, Yu.N.

    1988-01-01

    The classification spectrum results of random fluctuations in the IBR-2 energy pulse are presented. The work is performed for the application of the obtained results to the reactor diagnostics and the study of its noise uncontrolled states. For classification of the spectra the method of pattern recognition based upon the ISODATA heuristic algorithm is used. It is shown that a set of noise uncontrolled reactor states, registered during the reactor operation period at power of 0.4-2 MVt with the first variant of moving reflector (1983-1986) is formed into 4(5) most typical states. Each of the states corresponds to the general conditions of the reactor core cooling and provides the normal work of the moving reflector. However, these states differ in coolant flow, power level and peculiarities of the moving reflector rotation regime. One type of anomal power noise, connected with some disorder in the moving reflctor work, is isolated. This work also presents the possibility of control over the state of moving reflectors according to the change in the amplitude of power oscillations at some frequences. The reactor noise classification results can be used as the data bank for the IBR-2 reactor diagnostic system

  17. Research reactor FR2 - 20 years chemical and radiochemical measurements

    International Nuclear Information System (INIS)

    Feuerstein, H.; Graebner, H.; Oschinski, J.; Hoffmann, W.; Beyer, J.

    1986-09-01

    The FR2 has been a D 2 O cooled and moderated research reactor with a thermal output of 44 MW. It was in operation from 1961 to 1981. Because of the operating conditions of the reactor, only a small number of routine measurements were performed. For these however special techniques had to be developed. During the 20 years of operation a number of special events occured or have been observed, sometimes with very amazing results, e.g. the 'aceton effect'. This report describes the chemical and radiochemical conditions of the reactor systems, as well as the results of the surveilance work. Not described are measurements for the many experiments. The last chapter gives in a short form a description of the most unusual events and observations. (orig.) [de

  18. Venting krypton-85 from the Three Mile Island Unit 2 reactor building

    International Nuclear Information System (INIS)

    Burton, H.M.

    1981-01-01

    To permit the less restricted access to the reactor building necessary to maintain instrumentation and equipment, and to proceed towad the total decontamination of the facility, General Public Utilities, operators of the facility referred to hereafter as GPU, asked the United States Nuclear Regulatory Commission, or NRC, for permission to remove the 85 Kr from the reactor building by venting it to the environment. GPU supported their request with the Safety Analysis and Environmental Assessment Report on the proposed reactor building venting plan. On June 12, 1980, after seven months of licensing deliberations and numerous public hearings, the NRC granted GPU's request. The actual venting took place between June 28 and July 11, 1980. This report presents an overview of the detailed effort involved in the TMI-2 reactor building venting program. The findings reported here are condensed from a published report entitled TMI-2 Reactor Building Purge--Kr-85 Venting

  19. Research on economics and CO2 emission of magnetic and inertial fusion reactors

    International Nuclear Information System (INIS)

    Mori, Kenjiro; Yamazaki, Kozo; Oishi, Tetsutarou; Arimoto, Hideki; Shoji, Tatsuo

    2011-01-01

    An economical and environment-friendly fusion reactor system is needed for the realization of attractive power plants. Comparative system studies have been done for magnetic fusion energy (MFE) reactors, and been extended to include inertial fusion energy (IFE) reactors by Physics Engineering Cost (PEC) system code. In this study, we have evaluated both tokamak reactor (TR) and IFE reactor (IR). We clarify new scaling formulas for cost of electricity (COE) and CO 2 emission rate with respect to key design parameters. By the scaling formulas, it is clarified that the plant availability and operation year dependences are especially dominant for COE. On the other hand, the parameter dependences of CO 2 emission rate is rather weak than that of COE. This is because CO 2 emission percentage from manufacturing the fusion island is lower than COE percentage from that. Furthermore, the parameters dependences for IR are rather weak than those for TR. Because the CO 2 emission rate from manufacturing the laser system to be exchanged is very large in comparison with CO 2 emission rate from TR blanket exchanges. (author)

  20. Development of a TiO2-coated optical fiber reactor for water decontamination

    International Nuclear Information System (INIS)

    Danion, A.

    2004-09-01

    The objective of this study was to built and to study a photo-reactor composed by TiO 2 -coated optical fibers for water decontamination. The physico-chemical characteristics and the optical properties of the TiO 2 coating were first studied. Then, the influences of different parameters as the coating thickness, the coating length and the coating volume were investigated both on the light transmission in the TiO 2 - coated fiber and on the photo-catalytic activity of the fiber for a model compound (malic acid). The photo-catalytic degradation of malic acid was optimized using the experimental design methodology allowing to build a multi-fiber reactor comprising 57 optical fibers. The photo-degradation of malic acid was conducted in the multi-fiber reactor and it was demonstrated that the multi-fiber reactor was more efficient than the single-fiber reactor at the same fibers density. Finally, the multi-fiber reactor was applied to the photo-degradation of a fungicide, called fenamidone, and a degradation pathway was proposed. (author)

  1. Stability of narcissistic personality disorder: tracking categorical and dimensional rating systems over a two-year period.

    Science.gov (United States)

    Vater, Aline; Ritter, Kathrin; Strunz, Sandra; Ronningstam, Elsa F; Renneberg, Babette; Roepke, Stefan

    2014-07-01

    Personality disorders are characterized as temporally stable patterns of symptoms (APA, 2000). However, evidence on the stability of narcissistic personality disorder (NPD) is generally lacking. This study tracked the prevalence and remission rates of individual criteria for NPD over the course of 2 years. In addition, the stability of dimensional personality pathology in patients with NPD (assessed with the Dimensional Assessment of Personality Pathology, DAPP-BQ) was assessed over time. A sample of 96 patients with a diagnosis of NPD was recruited at baseline. Forty patients participated in the follow-up assessment 2 years later. Our results indicate a moderate remission rate (53%) for NPD as a categorical diagnosis. However, single NPD criteria differed in their prevalence and temporal stability, similar to findings for other personality disorders. Moreover, scores on dimensional subscales of the DAPP-BQ remained stable over time. Theoretical implications are discussed. (c) 2014 APA, all rights reserved.

  2. TMI-2 reactor-vessel head removal and damaged-core-removal planning

    International Nuclear Information System (INIS)

    Logan, J.A.; Hultman, C.W.; Lewis, T.J.

    1982-01-01

    A major milestone in the cleanup and recovery effort at TMI-2 will be the removal of the reactor vessel closure head, planum, and damaged core fuel material. The data collected during these operations will provide the nuclear power industry with valuable information on the effects of high-temperature-dissociated coolant on fuel cladding, fuel materials, fuel support structural materials, neutron absorber material, and other materials used in reactor structural support components and drive mechanisms. In addition, examination of these materials will also be used to determine accident time-temperature histories in various regions of the core. Procedures for removing the reactor vessel head and reactor core are presented

  3. 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)

  4. 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

  5. A theoretical analysis of methanol synthesis from CO2 and H2 in a ceramic membrane reactor

    NARCIS (Netherlands)

    Gallucci, F.; Basile, A.

    2007-01-01

    In this theoretical work the CO2 conversion into methanol in both a traditional reactor (TR) and a membrane reactor (MR) is considered. The purpose of this study was to investigate the possibility of increasing CO2 conversion into methanol with respect to a TR. A zeolite MR, able to combine

  6. Cronos 2: a neutronic simulation software for reactor core calculations

    International Nuclear Information System (INIS)

    Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M.

    1999-01-01

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  7. Set of rules SOR 2 reactor site criteria

    International Nuclear Information System (INIS)

    1976-06-01

    The purpose of this set of rules is to describe criteria which guide the Director in his evaluation of the suitability of proposed sites for stationary power and testing reactors subject to SOR 2. (B.G.)

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

    International Nuclear Information System (INIS)

    Liu Guisheng

    1995-01-01

    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 testings of CENDL-2 and ENDF/B-6. (4 tabs., 2 figs.)

  9. Market research companies and new product development tools

    NARCIS (Netherlands)

    Nijssen, E.J.; Frambach, R.T.

    1998-01-01

    This research investigates (1) the share of new product development (NPD) research services in market research (MR) companies’ turnover, (2) MR companies’ awareness and use of NPD tools and the modifications made to these NPD tools, and (3) MR company managers’ perceptions of the influence of client

  10. Market research companies and new product development tools

    NARCIS (Netherlands)

    Nijssen, Edwin J.; Frambach, Ruud T.

    1998-01-01

    This research investigates (1) the share of new product development (NPD) research services in market research (MR) companies' turnover, (2) MR companies' awareness and use of NPD tools and the modifications made to these NPD tools, and (3) MR company managers' perceptions of the influence of client

  11. Status of irradiation testing and PIE of MOX (Pu-containing) fuel

    International Nuclear Information System (INIS)

    Dimayuga, F.C.; Zhou, Y.N.; Ryz, M.A.

    1995-01-01

    This paper describes AECL's mixed oxide (MOX) fuel-irradiation and post-irradiation examination (PIE) program. Post-irradiation examination results of two major irradiation experiments involving several (U, Pu)O 2 fuel bundles are highlighted. One experiment involved bundles irradiated to burnups ranging fro 400 to 1200 MWh/kgHe in the Nuclear Power Demonstration (NPD) reactor. The other experiment consisted of several (U, Pu)O 2 bundles irradiated to burnups of up to 500 Mwh/kgHe in the National Research Universal (NRU) reactor. Results of these experiments demonstrate the excellent performance of CANDU MOX fuel. This paper also outlines the status of current MOX fuel irradiation tests, including the irradiation of various (U, Pu)O 2 bundles. The strategic importance of MOX fuel to CANDU fuel-cycle flexibility is discussed. (author)

  12. Groundwater Monitoring Plan for the Reactor Technology Complex Operable Unit 2-13

    International Nuclear Information System (INIS)

    Richard P. Wells

    2007-01-01

    This Groundwater Monitoring Plan describes the objectives, activities, and assessments that will be performed to support the on-going groundwater monitoring requirements at the Reactor Technology Complex, formerly the Test Reactor Area (TRA). The requirements for groundwater monitoring were stipulated in the Final Record of Decision for Test Reactor Area, Operable Unit 2-13, signed in December 1997. The monitoring requirements were modified by the First Five-Year Review Report for the Test Reactor Area, Operable Unit 2-13, at the Idaho National Engineering and Environmental Laboratory to focus on those contaminants of concern that warrant continued surveillance, including chromium, tritium, strontium-90, and cobalt-60. Based upon recommendations provided in the Annual Groundwater Monitoring Status Report for 2006, the groundwater monitoring frequency was reduced to annually from twice a year

  13. 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

  14. An experimental investigation of fission product release in SLOWPOKE-2 reactors - Data report

    International Nuclear Information System (INIS)

    Harnden, A.M.C.

    1995-09-01

    The results of an investigation into the release of fission products from SLOWPOKE-2 reactors fuelled with a highly-enriched uranium alloy core are detailed in Volume 1. This data report (Volume 2) contains plots of the activity concentrations of the fission products observed in the reactor container at the University of Toronto, Ecole Polytechnique and the Kanata Isotope Production Facility. Release rates from the reactor container water to the gas headspace are also included. (author)

  15. White organic light emitting diodes based on DCM dye sandwiched in 2-methyl-8-hydroxyquinolinolatolithium

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Amit [Center for Organic Electronics, Polymeric and Soft Materials Section, National Physical Laboratory (Council of Scientific and Industrial Research), Dr. K.S. Krishnan Road, New Delhi 110012 (India); Department of Chemistry, M.D. University, Rohtak, Haryana 124001 (India); Srivastava, Ritu, E-mail: ritu@mail.nplindia.ernet.i [Center for Organic Electronics, Polymeric and Soft Materials Section, National Physical Laboratory (Council of Scientific and Industrial Research), Dr. K.S. Krishnan Road, New Delhi 110012 (India); Bawa, Sukhwant S. [Center for Organic Electronics, Polymeric and Soft Materials Section, National Physical Laboratory (Council of Scientific and Industrial Research), Dr. K.S. Krishnan Road, New Delhi 110012 (India); Singh, Devender; Singh, Kapoor [Department of Chemistry, M.D. University, Rohtak, Haryana 124001 (India); Chauhan, Gayatri [Center for Organic Electronics, Polymeric and Soft Materials Section, National Physical Laboratory (Council of Scientific and Industrial Research), Dr. K.S. Krishnan Road, New Delhi 110012 (India); Singh, Ishwar [Department of Chemistry, M.D. University, Rohtak, Haryana 124001 (India); Kamalasanan, Modeeparampil N. [Center for Organic Electronics, Polymeric and Soft Materials Section, National Physical Laboratory (Council of Scientific and Industrial Research), Dr. K.S. Krishnan Road, New Delhi 110012 (India)

    2010-08-15

    Stable white electroluminescence (EL) has been achieved from organic LED, in which an ultrathin 4-(dicyanomethylene)-2-methyl-6-(p-dimethyl-aminostyryl)-4H-pyran (DCM) dye layer has been inserted in between two 2-methyl-8-hydroxyquinolinolatolithium [LiMeq] emitter layer and by optimizing the position of the DCM dye layer from the {alpha}-NPD/LiMeq interface. Electroluminescence spectra, current-voltage-luminescence (I-V-L) characteristics of the devices have been studied by changing the position of the dye layer. As the distance of DCM layer from {alpha}-NPD/LiMeq interface is increased, the intensity of host emission enhances rapidly. Introduction of thin layer of DCM in emissive layer increases the turn on voltage. The best Commission International de L' Eclairage (CIE) coordinates i.e. (0.32, 0.33) were obtained with device structure ITO/{alpha}-NPD(30 nm) /LiMeq(10 nm)/DCM(1 nm)/LiMeq(25 nm)/BCP(6 nm)/Alq{sub 3}(28 nm)/LiF(1 nm)/Al(100 nm). The EL spectrum covers the whole visible spectra range 400-700 nm. The color rendering index (CRI) for our best white light (Device 4) is 47.4. The device shows very good color stability in terms of CIE coordinates with voltages. The maximum luminescence 1240 cd/m{sup -2} has been achieved at 19 V.

  16. Safety assessments relating to the use of new fuels in research reactors: application to the case of FRM 2 reactor fuel

    International Nuclear Information System (INIS)

    Abou Yehia, H.; Bars, G.; Tran Dai

    2001-01-01

    After giving a brief reminder of the procedure applied in France for the licensing of the use of a new fuel type or design in a research reactor, we outline the main safety aspects associated with such a modification. Finally, by way of an example, we focus on the safety assessment relating to the IRIS irradiation device used in SILOE reactor, in particular for the qualification of the fuel dedicated to FRM II reactor of the Technical University of Munich. This qualification was carried out on a U 3 Si 2 fuel plate enriched to about 90 % in weight of 235 U and containing 1.5 g of uranium per cm 3 . The evaluation performed by the IPSN for GRS did not call into question the choice of U 3 Si 2 fuel plates for the FRM-II reactor. (authors)

  17. Thermal neutron flux distribution in ET-RR-2 reactor thermal column

    Directory of Open Access Journals (Sweden)

    Imam Mahmoud M.

    2002-01-01

    Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

  18. Apollo-L2, an advanced fuel tokamak reactor utilizing direct conversion

    International Nuclear Information System (INIS)

    Emmert, G.A.; Kulcinski, G.L.; Blanchard, J.P.; El-Guebaly, L.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.E.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Witt, R.J.

    1989-01-01

    A scoping study of a tokamak reactor fueled by a D- 3 He plasma is presented. The Apollo D- 3 He tokamak capitalizes on recent advances in high field magnets (20 T) and utilizes rectennas to convert the synchrotron radiation directly to electricity. The low neutron wall loading (0.1 MW/m 2 ) permits a first wall lasting the life of the plant and enables the reactor to be classified as inherently safe. The cost of electricity is less than that from a similar power level DT reactor. 10 refs., 1 fig., 4 tabs

  19. 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

  20. Rapid data acquisition from the safety system of the FRJ-2 reactor

    International Nuclear Information System (INIS)

    Inhoven, H.

    1980-06-01

    The central department for research reactors (ZFR) of the Juelich Nuclear Research Centre (KFA) is operating the reactors FRJ-1 (MERLIN) and FRJ-2 (DIDO) since 1962. In 1976, a Siemens 330 computer has been put into operation especially for the processing of data from the DIDO reactor, followed by another computer of the same type for the purpose of processing data from the ZFR department in general. The present report is a result of the work investigating 'Data acquisition and data processing in the FRJ-2' and primarily discusses the complex of 'fast analog and binary signals'. The activities in this field of work have been and still are mainly concerned with general problems encountered in adapting a currently 14-year-old reactor system to a digital computer, namely problems such as data decoupling in the safety system of the reactor, data acquisition using the CAMAC system, data transfer via an 'extended branch', data acquisition software as core-resident programs, temporary storage as common data, interpreting software as peripheral - storage - resident programs. (orig./WB) [de

  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. Estimation of power feedback parameters of pulse reactor IBR-2M on transients

    International Nuclear Information System (INIS)

    Pepyolyshev, Yu.N.; Popov, A.K.

    2013-01-01

    Parameters of the IBR-2M reactor power feedback (PFB) on a model of the reactor dynamics by mathematical treatment of two registered transients are estimated. Frequency characteristics and the pulse transient characteristics corresponding to these PFB parameters are calculated. PFB parameters received thus can be considered as their express tentative estimation as real measurements in this case occupy no more than 30 minutes. Total PFB is negative at 1 and 2 MW. At the received estimations of PFB parameters in a self-regulation mode it is possible to consider the stability margins of the IBR-2M reactor satisfactory

  4. Independent CO2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Task 2.50.05

    International Nuclear Information System (INIS)

    Stojic, M.; Pavicevic, M.

    1964-01-01

    This report contains the following volumes V and VI of the Project 'Independent CO 2 loop for cooling the samples irradiated in RA reactor vertical experimental channels': Design project of the dosimetry control system in the independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, and Safety report for the Independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels [sr

  5. Simulation of generation of new ideas for new product development and IT services

    Science.gov (United States)

    Nasiopoulos, Dimitrios K.; Sakas, Damianos P.; Vlachos, D. S.; Mavrogianni, Amanda

    2015-02-01

    This paper describes a dynamic model of the New Product Development (NPD) process. The model has been occurring from best practice noticed in our research conducted at a range of situations. The model contributes to determine and put an IT company's NPD activities into the frame of the overall NPD process[1]. It has been found to be a useful tool for organizing data on IT company's NPD activities without enforcement an excessively restrictive research methodology refers to the model of NPD. The framework, which strengthens the model, will help to promote a research of the methods undertaken within an IT company's NPD process, thus promoting understanding and improvement of the simulation process[2]. IT companies tested many techniques with several different practices designed to improve the validity and efficacy of their NPD process[3]. Supported by the model, this research examines how widely accepted stated tactics are and what impact these best tactics have on NPD performance. The main assumption of this study is that simulation of generation of new ideas[4] will lead to greater NPD effectiveness and more successful products in IT companies. With the model implementation, practices concern the implementation strategies of NPD (product selection, objectives, leadership, marketing strategy and customer satisfaction) are all more widely accepted than best practices related with controlling the application of NPD (process control, measurements, results). In linking simulation with impact, our results states product success depends on developing strong products and ensuring organizational emphasis, through proper project selection. Project activities strengthens both product and project success. IT products and services success also depends on monitoring the NPD procedure through project management and ensuring team consistency with group rewards. Sharing experiences between projects can positively influence the NPD process.

  6. PSA Level 2 activities for RBMK reactors

    International Nuclear Information System (INIS)

    Gubler, R.

    1998-01-01

    Probabilistic safety analyses (PSAs) of the boiling water graphite moderated pressure tube reactors (RBMKs) have been developed only recently and they are limited to Level 1. Activities at the IAEA were first motivated because of the difficulties to characterize core damage for RBMK reactors. Core damage probability is used in documents of the IAEA as a convenient single valued measure, for example for probabilistic safety criteria. The limited number of PSAs that have been completed for the RBMK reactors have shown that several special features of these channel type reactors necessitate revisiting of the characterization of core damage for these reactors. Furthermore, it has become increasingly evident that detailed deterministic analysis of DBAs and beyond design basis accidents reveal considerable insights into RBMK response to various accident conditions. These analyses can also help in better characterizing the outstanding phenomenological uncertainties, improved EOPs and AM strategies, including potential risk-beneficial accident negative backfits. The deterministic efforts should be focused first on elucidating accident progression processes and phenomena, and second on finding, qualifying and implementing procedures to minimize the risk of severe accident states The IAEA PSA procedures were mainly developed in New of vessel type LWRs, and would therefore require extensions to make them directly applicable. to channel type reactors. (author) (author)

  7. Studsvik's R2 reactor - Review of activities

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, Mikael; Tomani, Hans; Graeslund, Christian; Rundquist, Hans; Skoeld, Kurt [Studsvik Nuclear AB, Nykoeping (Sweden)

    1993-07-01

    A general description of the R2 reactor, its associated facilities and its history is given. The facilities and range of work are described for the following types of activities: fuel testing, materials testing, neutron transmutation doping of silicon, activation analysis, radioisotope production and basic research including thermal neutron scattering, nuclear chemistry and neutron capture radiography. (author)

  8. Decommissioning of reactor facilities (2). Required technology

    International Nuclear Information System (INIS)

    Yanagihara, Satoshi

    2014-01-01

    Decommissioning of reactor facilities was planned to perform progressive dismantling, decontamination and radioactive waste disposal with combination of required technology in a safe and economic way. This article outlined required technology for decommissioning as follows: (1) evaluation of kinds and amounts of residual radioactivity of reactor facilities with calculation and measurement, (2) decontamination technology of metal components and concrete structures so as to reduce worker's exposure and production of radioactive wastes during dismantling, (3) dismantling technology of metal components and concrete structures such as plasma arc cutting, band saw cutting and controlled demolition with mostly remote control operation, (3) radioactive waste disposal for volume reduction and reuse, and (4) project management of decommissioning for safe and rational work to secure reduction of worker's exposure and prevent the spreading of contamination. (T. Tanaka)

  9. Reproduction of the PSBR reactor with Exterminator-2; Reproduccion del reactor PSBR con exterminador-2

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1983-08-15

    To reproduce the reactor PSBR reported in (1), with the available version of the Exterminator-II in the ININ, they took the dimensions, composition specifications, effective sections of the different compositions (excepting those of the central thimble and of the moderator), the K{sub eff} and the factors of power (FP) for the different burners. Based on the comparison of the K{sub eff} and of the FP obtained with those reported the precision it is determined before in the reproduction of the reactor mentioned. (Author)

  10. Production of Sn-117m in the BR2 high-flux reactor.

    Science.gov (United States)

    Ponsard, B; Srivastava, S C; Mausner, L F; Russ Knapp, F F; Garland, M A; Mirzadeh, S

    2009-01-01

    The BR2 reactor is a 100MW(th) high-flux 'materials testing reactor', which produces a wide range of radioisotopes for various applications in nuclear medicine and industry. Tin-117m ((117m)Sn), a promising radionuclide for therapeutic applications, and its production have been validated in the BR2 reactor. In contrast to therapeutic beta emitters, (117m)Sn decays via isomeric transition with the emission of monoenergetic conversion electrons which are effective for metastatic bone pain palliation and radiosynovectomy with lesser damage to the bone marrow and the healthy tissues. Furthermore, the emitted gamma photons are ideal for imaging and dosimetry.

  11. 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

  12. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

  13. Comparative study between fluidized bed and fixed bed reactors in methane reforming with CO2 and O2 to produce syngas

    International Nuclear Information System (INIS)

    Jing Qiangshan; Lou Hui; Mo Liuye; Zheng Xiaoming

    2006-01-01

    Reforming of methane with carbon dioxide and oxygen was investigated over Ni/MgO-SiO 2 catalysts using fixed bed and fluidized bed reactors. The conversions of CH 4 and CO 2 in a fluidized bed reactor were close to thermodynamic equilibrium. The activity and stability of the catalyst in the fixed bed reactor were lower than that in the fluidized bed reactor due to carbon deposition and nickel sintering. TGA and TEM techniques were used to characterize the spent catalysts. The results showed that a lot of whisker carbon was found on the catalyst in the rear of the fixed bed reactor, and no deposited carbon was observed on the catalysts in the fluidized bed reactor after reaction. It is suggested that this phenomenon is related to a permanent circulation of catalyst particles between the oxygen rich and oxygen free zones. That is, fluidization of the catalysts in the fluidized bed reactor favors inhibiting deposited carbon and thermal uniformity in the reactor

  14. Benchmark testing of Canadol-2.1 for heavy water reactor

    International Nuclear Information System (INIS)

    Liu Ping

    1999-01-01

    The new version evaluated nuclear data library of ENDF-B 6.5 has been released recently. In order to compare the quality of evaluated nuclear data CENDL-2.1 with ENDF-B 6.5, it is necessary to do benchmarks testing for them. In this work, CENDL-2.1 and ENDF-B 6.5 were used to generated the WIMS-69 group library respectively, and benchmarks testing was done for the heavy water reactor, using WIMS5A code. It is obvious that data files of CENDL-2.1 is better than that of old WIMS library for the heavy water reactors calculations, and is in good agreement with those of ENDF-B 6.5

  15. Oxygen suppression in boiling water reactors. Phase 2. Annual report 1981, December 2, 1980-December 31, 1981

    International Nuclear Information System (INIS)

    Burley, E.L.

    1982-07-01

    A hydrogen addition test will be performed in the Dresden-2 reactor of Commonwealth Edison Company during 1982. Up to 2 ppM hydrogen will be added to and dissolved in the reactor feedwater to reverse the radiolysis reaction in the reactor core and suppress oxgen concentration in the primary coolant. At low oxygen levels the propensity of stressed and sensitized 304 stainless steel toward intergranular stress corrosion cracking is greatly reduced. The test will answer outstanding questions and uncertainties in the areas of water chemistry, equipment design and materials performance. Nine special sample facilities will be prepared in the primary coolant, main stream, feedwater/condensate, and offgas systems. Instrumentation will be available to measure hydrogen, oxygen, conductivity, pH, soluble and insoluble corrosion products, and electrochemical potentials. In addition, an autoclave in which confirming constant extension rate tests can be conducted in reactor water will be provided

  16. Characterization of fuel distributions in the Three-Mile Island Unit 2 (TMI-2) reactor system by neutron and gamma-ray dosimetry

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McNeece, J.P.; Kaiser, B.J.; McElroy, W.N.

    1984-04-01

    The resolution of technical issues generated by the accident at Three-Mile Island Unit 2 (TMI-2) will inevitably be of long range benefit. Determination of the fuel debris dispersal in the TMI-2 reactor system represents a major technical issue. In reactor recovery operations, such as for the safe handling and final disposal of TMI-2 waste, quantitative fuel assessments are being conducted throughout the reactor core and primary coolant system

  17. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    International Nuclear Information System (INIS)

    Bissani, M; O'Kelly, D S

    2006-01-01

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to provide color-enhanced gemstones but is

  18. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Bissani, M; O' Kelly, D S

    2006-05-08

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to

  19. analysis and implementation of reactor protection system circuits - case study Egypt's 2 nd research reactor-

    International Nuclear Information System (INIS)

    Elnokity, O.E.M.

    2006-01-01

    this work presents a way to design and implement the trip unit of a reactor protection system (RPS) using a field programmable gate arrays (FPGA). instead of the traditional embedded microprocessor based interface design method, a proposed tailor made FPGA based circuit is built to substitute the trip unit (TU), which is used in Egypt's 2 nd research reactor ETRR-2. the existing embedded system is built around the STD32 field computer bus which is used in industrial and process control applications. it is modular, rugged, reliable, and easy-to-use and is able to support a large mix of I/O cards and to easily change its configuration in the future. therefore, the same bus is still used in the proposed design. the state machine of this bus is designed based around its timing diagrams and implemented in VHDL to interface the designed TU circuit

  20. Attachment and mentalization in female patients with comorbid narcissistic and borderline personality disorder.

    Science.gov (United States)

    Diamond, Diana; Levy, Kenneth N; Clarkin, John F; Fischer-Kern, Melitta; Cain, Nicole M; Doering, Stephan; Hörz, Susanne; Buchheim, Anna

    2014-10-01

    We investigated attachment representations and the capacity for mentalization in a sample of adult female borderline patients with and without comorbid narcissistic personality disorder (NPD). Participants were 22 borderline patients diagnosed with comorbid NPD (NPD/BPD) and 129 BPD patients without NPD (BPD) from 2 randomized clinical trials. Attachment and mentalization were assessed on the Adult Attachment Interview (AAI; George, Kaplan, & Main, 1996). Results showed that as expected, compared with the BPD group, the NPD/BPD group was significantly more likely to be categorized as either dismissing or cannot classify on the AAI, whereas the BPD group was more likely to be classified as either preoccupied or unresolved for loss and abuse than was the NPD/BPD group. Both groups of patients scored low on mentalizing, and there were no significant differences between the groups, indicating that both NPD/BPD and BPD individuals showed deficits in this capacity. The clinical implications of the group differences in AAI classification are discussed with a focus on how understanding the attachment representations of NPD/BPD patients helps to illuminate their complex, contradictory mental states. (PsycINFO Database Record (c) 2014 APA, all rights reserved).

  1. Profiling pathological narcissism according to DSM-5 domains and traits: A study on consecutively admitted Italian psychotherapy patients.

    Science.gov (United States)

    Fossati, Andrea; Somma, Antonella; Borroni, Serena; Pincus, Aaron L; Markon, Kristian E; Krueger, Robert F

    2017-11-01

    [Correction Notice: An Erratum for this article was reported in Vol 29(11) of Psychological Assessment (see record 2016-56886-001). In the article, several values were reversed and the mean was misreported in Table 2. The corrected table is present in the erratum.] Pathological narcissism represents a clinically relevant, albeit controversial personality construct, with multiple conceptualizations that are operationalized by different measures. Even in the recently published Diagnostic and Statistical Manual for Mental Disorders-Fifth Edition (DSM-5), 2 different views of narcissistic personality disorder (NPD) are formulated (i.e., Section II and Section III). The DSM-5 Section III alternative PD model diagnosis of NPD is based on self and interpersonal dysfunction (Criterion A) and a profile of maladaptive personality traits (Criterion B), specifically elevated scores on Attention Seeking and Grandiosity. Given the diversity of conceptualizations of pathological narcissism, we evaluated the convergences and divergences in DSM-5 trait profiles characterizing multiple measures of narcissism in a clinical sample of 278 consecutively admitted Italian psychotherapy patients. Patients were administered the Italian versions of the Personality Inventory for DSM-5 (PID-5) and 4 measures of NPD, (a) the Narcissistic Personality Inventory (NPI); (b) the NPD scale of the Personality Diagnostic Questionnaire-4+; (c) the Structured Clinical Interview for Axis II Personality Disorders, Version 2.0 (SCID-II) as an observer-rated measure of NPD; and (d) the Pathological Narcissism Inventory (PNI). Multiple regression analyses showed that PID-5 traits explained from 13% to more than 60% of the variance in the different NPD measures. Attention Seeking was consistently associated with all measures of NPD, whereas Grandiosity was associated with some of the NPD measures. All measures of NPD were also significantly related to additional DSM-5 maladaptive traits. (PsycINFO Database

  2. Structural properties of the geometrically frustrated pyrochlore Tb2Ti2O7

    International Nuclear Information System (INIS)

    Han, Sang-Wook; Gardner, Jason S.; Booth, Corwin H.

    2004-01-01

    Although materials that exhibit nearest-neighbor-only antiferromagnetic interactions and geometrical frustration theoretically should not magnetically order in the absence of disorder, few such systems have been observed experimentally. One such system appears to be the pyrochlore Tb 2 Ti 2 O 7 . However, previous structural studies indicated that Tb 2 Ti 2 O 7 is an imperfect pyrochlore. To clarify the situation, we performed neutron powder diffraction (NPD) and x-ray absorption fine structure (XAFS) measurements on samples that were prepared identically to those that show no magnetic order. The NPD measurements show that the long-range structure of Tb 2 Ti 2 O 7 is well ordered with no structural transitions between 4.5 and 600 K. In particular, mean-squared displacements (u 2 's) for each site follow a Debye model with no offsets. No evidence for Tb/Ti site interchange was observed within an upper limit of 2%. Likewise, no excess or deficiency in the oxygen stoichiometry was observed, within an upper limit of 2% of the nominal pyrochlore value. Tb L III and Ti K-edge XAFS measurements from 20-300 K similarly indicate a well-ordered local structure. Other aspects of the structure are considered. We conclude that Tb 2 Ti 2 O 7 has, within experimental error, an ideal, disorder-free pyrochlore lattice, thereby allowing the system to remain in a dynamic, frustrated spin state to the lowest observed temperatures

  3. Determination of ifosfamide, 2-and 3-dechloroethyifosfamide using gas chromatography with nitrogen-phosphorus or mass spectrometry detection

    NARCIS (Netherlands)

    Kerbusch, T; Jeuken, MJ; Derraz, J; van Putten, JWG; Huitema, ADR; Beijnen, JH

    2000-01-01

    A comparison was made between methods for determining ifosfamide (IF), 2- (2DCE) and 3-dechloroethylifosfamide (3DCE) using gas chromatography with nitrogen-phosphorus detection (GC-NPD) versus positive ion electron-impact ion-trap mass spectrometry (GC-MS'). Sample pretreatment involved

  4. General outline of the operation and utilization of the BR2 reactor

    International Nuclear Information System (INIS)

    Baugnet, J.M.; Leonard, F.; Gandolfo, J.M.; Lenders, H.

    1978-01-01

    The BR2 reactor is a high-flux material testing reactor of the thermal heterogeneous type. The fuel is 93% 235 U enriched uranium in the form of plates clad in aluminium. The moderator consists of beryllium and light water, the water being pressurized (12.5kg/cm 2 )and acting also as coolant. The pressure vessel is of aluminium, and is placed in a pool of demineralized water. One should stress the following main features of the design: the experimental channels are skew, the tube bundle presenting the form of a hyperboloid of revolution (see figure 1)-this gives easy access at the top and bottom reactor covers allowing complex instrumented devices, while maintaining a very high neutron flux at the core; great flexibilty of utilization, due to the fact that it is possible to adapt the core configuration to the experimental loading as the fissile charge can be centred on different experimental channels; although BR2 is a thermal reactor, it is possible to achieve neutron spectra very similar to those obtained in a fast reactor, either by the use of absorbing screens or by the use of fissile material within the experimental device; five 200mm diameter channels are available for loading large experimental irradiation devices, as in-pile sodium, gas or water loops. (author)

  5. Benchmark testing of CENDL-2 for U-fuel thermal reactors

    International Nuclear Information System (INIS)

    Zhang Baocheng; Liu Guisheng; Liu Ping

    1995-01-01

    Based on CENDL-2, NJOY-WIMS code system was used to generate 69-group constants, and do benchmark testing for TRX-1,2; BAPL-UO-2-1,2,3; ZEEP-1,2,3. All the results proved that CENDL-2 is reliable for thermal reactor calculations. (3 tabs.)

  6. Narcissistic Personality Disorder and suicidal behavior in mood disorders.

    Science.gov (United States)

    Coleman, Daniel; Lawrence, Ryan; Parekh, Amrita; Galfalvy, Hanga; Blasco-Fontecilla, Hilario; Brent, David A; Mann, J John; Baca-Garcia, Enrique; Oquendo, Maria A

    2017-02-01

    The relationship of Narcissistic Personality Disorder (NPD) to suicidal behavior is understudied. The modest body of existing research suggests that NPD is protective against low-lethality suicide attempts, but is associated with high lethality attempts. Mood-disordered patients (N = 657) received structured interviews including Axis I and II diagnosis and standardized clinical measures. Following chi-square and t-tests, a logistical regression model was constructed to identify predictors of suicide attempt. While there was no bivariate relationship of NPD on suicide attempt, in the logistic regression patients with NPD were 2.4 times less likely to make a suicide attempt (OR = 0.41; 95% CI = 0.19 - 0.88; p disorder, and to have high aggression and hostility scores. Limitations include that the sample consists of only mood-disordered patients, a modest sample size of NPD, and the data are cross-sectional. The multivariate protective effect of NPD on suicide attempt is consistent with most previous research. The lower impulsivity of NPD patients and less severe personality pathology relative to other personality disorders may contribute to this effect. No relationship of NPD to attempt lethality was found, contradicting other research, but perhaps reflecting differences between study samples. Future studies should oversample NPD patients and include suicide death as an outcome. Clinical implications include discussion of individualized suicide risk assessment with NPD patients. Copyright © 2016 Elsevier Ltd. All rights reserved.

  7. Shadow corrosion testing in the INCA facility in the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Nystrand, A.C.; Lassing, A.

    1999-01-01

    Shadow corrosion is a phenomenon which occurs when zirconium alloys are in contact with or in proximity to other metallic objects in a boiling water reactor environment (BWR, RBMK, SGHWR etc.). An enhanced corrosion occurs on the zirconium alloy with the appearance of a 'shadow' of the metallic object. The magnitude of the shadow corrosion can be significant, and is potentially limiting for the lifetime of certain zirconium alloy components in BWRs and other reactors with a similar water chemistry. In order to evaluate the suitability of the In-Core Autoclave (INCA) in the Studsvik R2 materials testing reactor as an experimental facility for studying shadow corrosion, a demonstration test has been performed. A number of test specimens consisting of Zircaloy-2 tubing in contact with Inconel were exposed in an oxidising water chemistry. Some of the specimens were placed within the reactor core and some above the core. The conclusion of this experiment after post irradiation examination is that it is possible to use the INCA facility in the Studsvik R2 reactor to develop a significant level of shadow corrosion after only 800 hours of irradiation. (author)

  8. The 5th surveillance testing for Kori unit 2 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

  9. Digital, remote control system for a 2-MW research reactor

    International Nuclear Information System (INIS)

    Battle, R.E.; Corbett, G.K.

    1988-01-01

    A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs

  10. Narcissistic personality disorder: an integrative review of recent empirical data and current definitions.

    Science.gov (United States)

    Roepke, Stefan; Vater, Aline

    2014-05-01

    Although concepts of pathological narcissism are as old as psychology and psychiatry itself, only a small number of clinical studies are based on the criteria for narcissistic personality disorder (NPD), as defined in the Diagnostic and Statistical Manuals of Mental Disorders (DSM). As a result, NPD appears to be one of the most controversially discussed nosological entities in psychiatry. Whereas the majority of empirical studies used self or other ratings of NPD criteria to address issues of reliability and validity of the diagnostic category (i.e., internal consistency, factor structure, discriminant validity), only recent research has applied experimental designs to investigate specific features of NPD (e.g., self-esteem, empathy, shame). The aim of this review is to summarize available empirical data on NPD and relate these findings to current definitions of NPD (according to the DSM-5, [1]). In order to do so, this review follows the five steps to establishing diagnostic validity proposed by Robins and Guze [2], i.e., (1) clinical description, (2) laboratory studies, (3) delimitation from other disorders, (4) family studies, and (5) follow up studies. Finally, this review suggests pathways for future research that may assist further nosological evaluation of NPD and contribute to the overall goal, the improvement of treatment for patients.

  11. A nodal Grean's function method of reactor core fuel management code, NGCFM2D

    International Nuclear Information System (INIS)

    Li Dongsheng; Yao Dong.

    1987-01-01

    This paper presents the mathematical model and program structure of the nodal Green's function method of reactor core fuel management code, NGCFM2D. Computing results of some reactor cores by NGCFM2D are analysed and compared with other codes

  12. RHTF 2, a 1200 MWe high temperature reactor

    International Nuclear Information System (INIS)

    Brisbois, Jacques

    1978-01-01

    After having adapted to French conditions the 1160 MWe G.A.C. reactor, Commissariat a l'Energie Atomique and French Industry have decided to design an High Temperature Reactor 1200 MWe based on the G.A.C. technology and taking into account the point of view of Electricite de France and the experience of C.E.A. and industry on the gas cooled reactor technology. The main objective of this work is to produce a reactor design having a low technical risk, good operability, with an emphasis on the safety aspects easing the licensing problems

  13. Turkey's regulatory plans for high enriched to low enriched conversion of TR-2 reactor core

    International Nuclear Information System (INIS)

    Guelol Oezdere, Oya

    2003-01-01

    Turkey is a developing country and has three nuclear facilities two of which are research reactors and one pilot fuel production plant. One of the two research reactors is TR-2 which is located in Cekmece site in Istanbul. TR-2 Reactor's core is composed of both high enriched and low enriched fuel and from high enriched to low enriched core conversion project will take place in year 2005. This paper presents the plans for drafting regulations on the safety analysis report updates for high enriched to low enriched core conversion of TR-2 reactor, the present regulatory structure of Turkey and licensing activities of nuclear facilities. (author)

  14. Techno-economic assessment of membrane assisted fluidized bed reactors for pure H_2 production with CO_2 capture

    International Nuclear Information System (INIS)

    Spallina, V.; Pandolfo, D.; Battistella, A.; Romano, M.C.; Van Sint Annaland, M.; Gallucci, F.

    2016-01-01

    Highlights: • Membrane reactors improve the overall efficiency of H_2 production up to 20%. • Respect to conventional reforming, the H_2 yield increases from 12% to 20%. • The COH is reduced of at least 220% using membrane reactors. • FBMR capture 72% of CO_2 with a specific cost of 8 eur/tonn_C_O_2_. • MA-CLR can reach 90% of CO_2 avoided with same cost of FTR. - Abstract: This paper addresses the techno-economic assessment of two membrane-based technologies for H_2 production from natural gas, fully integrated with CO_2 capture. In the first configuration, a fluidized bed membrane reactor (FBMR) is integrated in the H_2 plant: the natural gas reacts with steam in the catalytic bed and H_2 is simultaneously separated using Pd-based membranes, and the heat of reaction is provided to the system by feeding air as reactive sweep gas in part of the membranes and by burning part of the permeated H_2 (in order to avoid CO_2 emissions for heat supply). In the second system, named membrane assisted chemical looping reforming (MA-CLR), natural gas is converted in the fuel rector by reaction with steam and an oxygen carrier (chemical looping reforming), and the produced H_2 permeates through the membranes. The oxygen carrier is re-oxidized in a separate air reactor with air, which also provides the heat required for the endothermic reactions in the fuel reactor. The plants are optimized by varying the operating conditions of the reactors such as temperature, pressures (both at feed and permeate side), steam-to-carbon ratio and the heat recovery configuration. The plant design is carried out using Aspen Simulation, while the novel reactor concepts have been designed and their performance have been studied with a dedicated phenomenological model in Matlab. Both configurations have been designed and compared with reference technologies for H_2 production based on conventional fired tubular reforming (FTR) with and without CO_2 capture. The results of the analysis show

  15. Reactor building integrity testing: A novel approach at Gentilly 2 - principles and methodology

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1991-01-01

    In 1987, Hydro-Quebec embarked on an ambitious development program to provide the Gentilly 2 nuclear power station with an effective, yet practical reactor building Integrity Test. The Gentilly 2 Integrity Test employs an innovative approach based on the reference volume concept. It is identified as the Temperature Compensation Method (TCM) System. This configuration has been demonstrated at both high and low test pressure and has achieved extraordinary precision in the leak rate measurement. The Gentilly 2 design allows the Integrity Test to be performed at a nominal 3 kPa(g) test pressure during an (11) hour period with the reactor at full power. The reactor building Pressure Test by comparison, is typically performed at high pressure 124 kPa(g)) in a 7 day window during an annual outage. The Integrity Test was developed with the goal of demonstrating containment availability. Specifically it was purported to detect a leak or hole in the 'bottled-up' reactor building greater in magnitude than an equivalent pipe of 25 mm diameter. However it is considered feasible that the high precision of the Gentilly 2 TCM System Integrity Test and a stable reactor building leak characteristic will constitute sufficient grounds for the reduction of the Pressure Test frequency. It is noted that only the TCM System has, to this date, allowed a relevant determination of the reactor building leak rate at a nominal test pressure of 3 kPa(g). Classical method tests at low pressure have lead to inconclusive results due to the high lack of precision

  16. Maximum credible accident analysis for TR-2 reactor conceptual design

    International Nuclear Information System (INIS)

    Manopulo, E.

    1981-01-01

    A new reactor, TR-2, of 5 MW, designed in cooperation with CEN/GRENOBLE is under construction in the open pool of TR-1 reactor of 1 MW set up by AMF atomics at the Cekmece Nuclear Research and Training Center. In this report the fission product inventory and doses released after the maximum credible accident have been studied. The diffusion of the gaseous fission products to the environment and the potential radiation risks to the population have been evaluated

  17. UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

    Energy Technology Data Exchange (ETDEWEB)

    Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Aji, Indarta K. [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

    2016-03-11

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

  18. The Chernobyl reactor accident. Pt. 1 and 2

    International Nuclear Information System (INIS)

    1986-06-01

    The report first summarizes the available information on the various incidents of the whole accident scenario, and combines the information to present a first general outline and a basis for appraisal. The most significant incidents reported, namely power excursion, core meltdown, and fire, are discussed with a view to the reactor design and safety of reactors installed in the FRG. The main differences and advantages of German reactor designs are shown, as e.g.: Power excursions are mastered by inherent physical conditions; far better redundancy of engineered safety systems; enclosure of the complete reactor cooling system in a pressure-retaining steel containment; reactor buildings being made of reinforced concrete. The second part of the report deals with the radiological effects to be expected for our country. Data are given on the varying radiological exposure of the different regions. The fate and uptake of radioactivity in the human body are discussed. The conclusion drawn from the data presented is that the individual exposure due to the reactor accident will remain within the variations and limits of natural radioactivity and effects. (orig./HP) [de

  19. Synthesis of the IRSN report related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor. Referral of the Permanent Group of Experts for nuclear reactors (GPR), examination of probabilistic level-2 safety studies (EPS 2) and severe accidents (AG) of the Flamanville reactor nr 3. Opinion related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor (FA3). Electronuclear reactors - EDF - Flamanville 3 EPR reactor. Severe accidents and probabilistic level 2 studies

    International Nuclear Information System (INIS)

    2015-01-01

    This document gathers several documents. The first one recalls the main arrangements implemented on the FA3 EPR reactor regarding accidents with core fusion, reports the analysis made by the IRSN about the sizing of these arrangements to reach a controlled status of the installation after a severe accident, regarding the probabilistic level-2 safety assessment, regarding the radiological impact of a severe accident on the population and on the environment, regarding those aimed at facing a total and long duration loss of electric power sources and cold sources, and about the situation of the reactor with respect to WENRA positions on severe accidents for new reactors. The second document is a letter sent by the ASN to the Permanent Group of Experts for nuclear reactors (GPR) to address probabilistic level-2 safety studies (EPS2) and severe accidents for the Flamanville 3 reactor. The third one reports the opinion of the GPR on these both issues and proposes a set of recommendations. The next document is a letter sent by the ASN to the Flamanville 3 project manager at EDF which recalls the related objectives, the ASN opinion on the implemented arrangements for severe accidents (de-pressurization of the primary circuit, management of hydrogen-related risks, corium recovery and cooling outside the vessel, limitation of vapour explosion risks outside the vessel, heat evacuation system, containment enclosure, management of the risk of a return to criticality), to face a total and long duration loss of electricity sources and cold sources, and other aspects addressed in the IRSN analysis. Requests and remarks formulated by the ASN are provided in an appendix to this last document

  20. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  1. 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.)

  2. A Conceptual Study on a Supercritical CO_2-cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Yu, Hwanyeal; Hartanto, Donny; Kim, Yonghee

    2014-01-01

    A Micro Modular Reactor (MMR) using Supercritical-CO_2 (S-CO_2) as coolant has been investigated from the neutronics perspective. The MMR is designed to be transportable so it can reach the remote areas. The thermal power of the reactor is 36.2 M Wth. The size of the active core is limited to 1.2 m length and 93.16 cm width. The size of whole core is 2.8 m length and 166.9 cm width. The reactor lifetime design target is 20 years. To maximize the fuel volume fraction in the core, high density uranium nitride UN"1"5 was used. The PbO/MgO reflector was also utilized to improve the neutron economy. The S-CO_2 is chosen as the coolant because it offers a higher thermal efficiency. In this study, neutronics calculations and depletion using McCARD Monte Carlo code has been done to determine the lifetime and behavior of the core. Several important safety parameters such as Control Rod worth, Doppler reactivity coefficients and coolant void reactivity coefficient have also been analyzed. (author)

  3. Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Coles, G.A.; McKay, S.L.

    1995-06-01

    A multilateral team recently completed a full-scope Level 1 Probabilistic Safety Assessment (PSA) on the Ignalina Unit No. 2 reactor plant in Lithuania. This allows comparison of results to those of the PSA for the U.S. Department of Energy's (DOE) N Reactor. The N Reactor, although unique as a Western design, has similarities to Eastern European and Soviet graphite block reactors

  4. CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

    International Nuclear Information System (INIS)

    Kotas, J.F.; Stroh, K.R.

    1983-01-01

    The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power

  5. Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

    2010-01-01

    In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

  6. Preliminary Design of S-CO2 Brayton Cycle for KAIST Micro Modular Reactor

    International Nuclear Information System (INIS)

    Kim, Seong Gu; Kim, Min Gil; Bae, Seong Jun; Lee, Jeong Ik

    2013-01-01

    This paper suggests a complete modular reactor with an innovative concept of reactor cooling by using a supercritical carbon dioxide directly. Authors propose the supercritical CO 2 Brayton cycle (S-CO 2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the core and PCU in one vessel for the full modularization. This study suggests a conceptual design of small modular reactor including PCU which is named as KAIST Micro Modular Reactor (MMR). As a part of ongoing research of conceptual design of KAIST MMR, preliminary design of power generation cycle was performed in this study. Since the targets of MMR are full modularization of a reactor system with S-CO 2 coolant, authors selected a simple recuperated S-CO 2 Brayton cycle as a power conversion system for KAIST MMR. The size of components of the S-CO 2 cycle is much smaller than existing helium Brayton cycle and steam Rankine cycle, and whole power conversion system can be contained with core and safety system in one containment vessel. From the investigation of the power conversion cycle, recompressing recuperated cycle showed higher efficiency than the simple recuperated cycle. However the volume of heat exchanger for recompressing cycle is too large so more space will be occupied by heat exchanger in the recompressing cycle than the simple recuperated cycle. Thus, authors consider that the simple recuperated cycle is more suitable for MMR. More research for the KAIST MMR will be followed in the future and detailed information of reactor core and safety system will be developed down the road. More refined cycle layout and design of turbomachinery and heat exchanger will be performed in the future study

  7. Severe accident analysis for level 2 PSA of SMART reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jin Yong; Lee, Jeong Hun; Kim, Jong Uk; Yoo, Tae Geun; Chung, Soon Il; Kim, Min Gi [FNC Technology Co., Seoul (Korea, Republic of)

    2010-12-15

    The objectives of this study are to produce data for level 2 PSA and evaluation results of severe accident by analyzing severe accident sequence of transient events, producing fault tree of containment systems and evaluating direct containment heating of the SMART. In this project, severe accident analysis results were produced for general transient, loss of feedwater, station blackout, and steam line break events, and based on the results, design safety of SMART was verified. Also, direct containment heating phenomenon of the SMART was evaluated using TCE methodology. For level 2 PSA, fault tree of the containment isolation system, reactor cavity flooding system, plant chilled water system, and reactor containment building HVAC system was produced and analyzed

  8. Recent advances in the utilization and the irradiation technology of the refurbished BR2 reactor

    International Nuclear Information System (INIS)

    Dekeyser, J.; Benoit, P.; Decloedt, C.; Pouleur, Y.; Verwimp, A.; Weber, M.; Vankeerberghen, M.; Ponsard, B.

    1999-01-01

    Operation and utilization of the materials testing reactor BR2 at the Belgian Nuclear Research Centre (SCK·CEN) has since its start in 1963 always followed closely the needs and developments of nuclear technology. In particular, a multitude of irradiation experiments have been carried out for most types of nuclear power reactors, existing or under design. Since the early 1990s and increased focus was directed towards more specific irradiation testing needs for light water reactor fuels and materials, although other areas of utilization continued as well (e.g. fusion reactor materials, safety research, ...), including also the growing activities of radioisotope production and silicon doping. An important milestone was the decision in 1994 to implement a comprehensive refurbishment programme for the BR2 reactor and plant installations. The scope of this programme comprised very substantial studies and hardware interventions, which have been completed in early 1997 within planning and budget. Directly connected to this strategic decision for reactor refurbishment was the reinforcement of our efforts to requalify and upgrade the existing irradiation facilities and to develop advanced devices in BR2 to support emerging programs in the following fields: - LWR pressure vessel steel, - LWR irradiation assisted stress corrosion cracking (IASCC), - reliability and safety of high-burnup LWR fuel, - fusion reactor materials and blanket components, - fast neutron reactor fuels and actinide burning, - extension and diversification of radioisotope production. The paper highlights these advances in the areas of BR2 utilisation and the ongoing development activities for the required new generation of irradiations devices. (author)

  9. An improved thermal-hydraulic modeling of the Jules Horowitz Reactor using the CATHARE2 system code

    Energy Technology Data Exchange (ETDEWEB)

    Pegonen, R., E-mail: pegonen@kth.se [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden); Bourdon, S.; Gonnier, C. [CEA, DEN, DER, SRJH, CEA Cadarache, 13108 Saint-Paul-lez-Durance Cedex (France); Anglart, H. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden)

    2017-01-15

    Highlights: • An improved thermal-hydraulic modeling of the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during loss of flow accident. • The heat exchanger approach gives more realistic and less conservative results. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support current and future nuclear reactor designs. The reactor is under construction at the CEA Cadarache research center in southern France and is expected to achieve first criticality at the end of this decade. This paper presents an improved thermal-hydraulic modeling of the reactor using solely CATHARE2 system code. Up to now, the CATHARE2 code was simulating the full reactor with a simplified approach for the core and the boundary conditions were transferred into the three-dimensional FLICA4 core simulation. A new more realistic methodology is utilized to analyze the thermal-hydraulic simulation of the reactor during a loss of flow accident.

  10. 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

  11. Proceedings of 2. Yugoslav symposium on reactor physics, Part 2, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 2, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-07-01

    This Volume 2 of the Proceedings of 2. Yugoslav symposium on reactor physics includes eight papers dealing with the following topics: method for measuring high anti reactivities of a reactor system; integration method for thermal reaction rate calculation; Determination of initial core configuration for BHWR-200 MWe; safety shutdowns and failures of the RA reactor equipment; determining the reactivity of absorption rods; measurements of thermal and fast neutron fluxes at the TRIGA reactor and other measurements during operation of the TRIGA reactor; mathematical modelling of the reactor safety; review of problems and methods for radiation risk assessment in the environment of a nuclear power plant.

  12. Fuel channel design improvements for large CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Villamagna, A; Price, E G; Field, G J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    From the initial designs used in NPD and Douglas point reactors, the CANDU fuel channel and its components have undergone considerable development. Two major designs have evolved: the Pickering/CANDU 6 design which has 12 fuel bundles in the core and where the new fuel is inserted into the inlet end, and the Bruce/Darlington design which has 13 bundles in the channel and where new fuel is inserted into the outlet end. In the development of a single unit CANDU reactor of the size of a Bruce or Darlington unit which would use a Darlington design calandria, the decision has been made to use the CANDU 6 fuel channel rather than the Darlington design. The CANDU 6 channel has provided excellent performance and will not encounter the degree of maintenance required for the Bruce/Darlington design. The channel design in turn influences the fuelling machine/fuel handling concepts required. The changes to the CANDU 6 fuel channel design to incorporate it in the large unit are small. In fact, the changes that are proposed relate to the desire to increase margins between pressure tube properties and design conditions or ameliorate the consequences of postulated accident conditions, rather than necessary adaptation to the larger unit. Better properties have been achieved in the pressure tube material resulting from alloy development program over the past 10 years. Pressure tubes can now he made with very low hydrogen concentrations so that the hydrogen picked up as deuterium will not exceed the terminal solid solubility for the in-core region in 30 years. The improvements in metal chemistry allow the production of high toughness tubes that retain a high level of toughness during service. A small increase in wall thickness will reduce the dimensional changes without significantly affecting burnup. Changes to increase safety margins from postulated accidents are concentrated on containing the consequences of pressure tube damage. The changes are concentrated on the calandria tube

  13. Problems of nuclear reactor safety. Vol. 2

    International Nuclear Information System (INIS)

    Goncharov, L.A.

    1995-01-01

    Theses of proceedings of the 9 Topical Meeting on problems of nuclear power plant safety are presented. Reports include results of neutron-physical experiments carried out for reactor safety justification. Concepts of advanced reactors with improved safety are considered. Results of researches on fuel cycles are given too

  14. The Effect Of Beryllium Interaction With Fast Neutrons On the Reactivity Of ETRR-2 Research Reactor

    International Nuclear Information System (INIS)

    Aziz, M.; El Messiry, A.M.

    2000-01-01

    The effect of beryllium interactions with fast neutrons is studied for Etrr 2 research reactors. Isotope build up inside beryllium blocks is calculated under different irradiation times. a new model for the Etrr 2 research reactor is designed using MCNP code to calculate the reactivity and flux change of the reactor due to beryllium poison

  15. Economics and utilization of thorium in nuclear reactors. Technical annexes 1 and 2

    International Nuclear Information System (INIS)

    1978-05-01

    An assessment of the impact of utilizing the 233 U/thorium fuel cycle in the U.S. nuclear economy is strongly dependent upon several decisions involving nuclear energy policy. These decisions include: (1) to recycle or not recycle fissile material; (2) if fissile material is recycled, to recycle plutonium, 233 U, or both; and (3) to deploy or not to deploy advanced reactor designs such as Fast Breeder Reactors (FBR's), High Temperature Gas Reactors (HTGR's), and Canadian Deuterium Uranium Reactors (CANDU's). This report examines the role of thorium in the context of the above policy decisions while focusing special attention on economics and resource utilization

  16. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    International Nuclear Information System (INIS)

    Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Pal, S.; Mishra, G.P.

    2012-01-01

    CERMET fuel with either PuO 2 or enriched UO 2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO 2 compositions.

  17. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    Science.gov (United States)

    Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

    2012-08-01

    CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

  18. Utilization of the SLOWPOKE-2 research reactor

    International Nuclear Information System (INIS)

    Lalor, G.C.

    2001-01-01

    SLOWPOKEs are typically low power research reactors that have a limited number of applications. However, a significant range of NAA can be performed with such reactors. This paper describes a SLOWPOKE-based NAA program that is performing a valuable series of studies in Jamaica, including geological mapping and pollution assessment. (author)

  19. Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

  20. Measurements at the RA Reactor related to the VISA-2 project - Part 1, Start-up of the RA reactor and measurement of new RA reactor core parameters

    International Nuclear Information System (INIS)

    Markovic, H.

    1962-07-01

    The objective of the measurements was determining the neutron flux in the RA reactor core. Since the number of fuel channels is increased from 56 to 68 within the VISA-2 project, it was necessary to attain criticality of the RA reactor and measure the neutron flux properties. The 'program of RA reactor start-up' has been prepared separately and it is included in this report. Measurements were divided in two phases. First phase was measuring of the neutron flux after the criticality was achieved but at zero power. During phase two measurements were repeated at several power levels, at equilibrium xenon poisoning. This report includes experimental data of flux distributions and absolute values of the thermal and fast neutron flux in the RA reactor experimental channels and values of cadmium ratio for determining the neutron epithermal flux. Data related to calibration of regulatory rods for cold un poisoned core are included [sr

  1. Reactor limitation system improves the safety and availability of the Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

    Souza Mendes, J.E. de

    1987-01-01

    Beyond the classic Reactor Protection System and Reactor Control System, nuclear plant Angra 2 has a third system called Reactor Limitation System which combines the intelligence features of the control systems with the high reliability of the protection systems. In determined events, which are not controlled by the control system (e.g.: load rejection, failure of one main reactor coolant pump), the Reactor Limitation System actuates automatically in order to lead the plant to a safe operating condition and so it avoids the actuation of the Reactor Protection System and consequently the reactor trip. This increases safety and availability of the plant and reduces component stresses. After the safe operating condition is reached, the process guidance automatically returns to the control systems. (Author) [pt

  2. CO_2 capture with solid sorbent: CFD model of an innovative reactor concept

    International Nuclear Information System (INIS)

    Barelli, L.; Bidini, G.; Gallorini, F.

    2016-01-01

    Highlights: • A new reactor solution based on rotating fixed beds was presented. • The preliminary design of the reactor was approached. • A CFD model of the reactor, including CO_2 capture kinetic, was developed. • The CFD model is validated with experimental results. • Sorbent exploitation increasing is possible thanks to the new reactor. - Abstract: In future decarbonization scenarios, CCS with particular reference to post-combustion technologies will be an important option also for energy intensive industries. Nevertheless, today CCS systems are rarely installed due to high energy and cost penalties of current technology based on chemical scrubbing with amine solvent. Therefore, innovative solutions based on new/optimized solvents, sorbents, membranes and new process designs, are R&D priorities. Regarding the CO_2 capture through solid sorbents, a new reactor solution based on rotating fixed beds is presented in this paper. In order to design the innovative system, a suitable CFD model was developed considering also the kinetic capture process. The model was validated with experimental results obtained by the authors in previous research activities, showing a potential reduction of energy penalties respect to current technologies. In the future, the model will be used to identify the control logic of the innovative reactor in order to verify improvements in terms of sorbent exploitation and reduction of system energy consumption.

  3. MULTI-LOOP CONTROL DESIGN IN MULTIVARIABLE (2X2 CONTINUOUS STIRRED TANK REACTOR

    Directory of Open Access Journals (Sweden)

    Abdul Wahid

    2015-06-01

    Full Text Available With this study, the design and tuning of multi-loop for multivariable (2x2 CSTR will be made in order to achieve optimum CSTR control performance. This study used Bequette model reactor and MATLAB software and is expected to be able to cope with disturbances in the reactor so that the reactor system is able to stabilize quickly despite the distractions. In this study, the design will be made using multi-loop approach, along with PI controller as the next step. Then, BLT and auto-tune tuning method will be used in PI controller and given disturbances to both of tuning method. The controller performances are then compared. Results of the study are then analyzed for discussions and conclusions. Results from this study have shown that in terms of disturbance rejection, BLT is better than auto-tune based on comparison between both of controller performances. For IAE for the case of temperature, BLT is 30% better than auto-tune, but it is almost the same for the case of concentration. For settling time for the case of concentration, BLT is 30% better than auto-tune, and for the case of temperature, BLT is 18% better than auto-tune. For rise time for the case of concentration and temperature, BLT is 30% better than auto-tune.

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

    International Nuclear Information System (INIS)

    1985-04-01

    The Operating Reactors Licensing Actions 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. This summary report is published primarily for internal NRC use in managing the Operating Reactors Licensing Actions Program

  5. Tests of Neutron Spectrum Calculations with the Help of Foil Measurements in a D{sub 2}O and in an H{sub 2}O-Moderated Reactor and in Reactor Shields of Concrete an Iron

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, R; Aalto, E

    1964-09-15

    Foil measurements covering the fast, epithermal and thermal neutron energy regions have been made in the centre of the Swedish D{sub 2}O-moderated reactor R1, in the pool reactor R2-0, and in different positions in reactor shields of iron, magnetite concrete and ordinary concrete. Neutron spectra have also been calculated for most of these positions, often with the help of a numerical integration of the Boltzmann equation. The measurements and the calculated spectra are presented.

  6. Structural properties of the geometrically frustrated pyrochlore Tb2Ti2O7

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sang-Wook; Gardner, Jason S.; Booth, Corwin H.

    2004-06-14

    Although materials that exhibit nearest-neighbor-only antiferromagnetic interactions and geometrical frustration theoretically should not magnetically order in the absence of disorder, few such systems have been observed experimentally. One such system appears to be the pyrochlore Tb{sub 2}Ti{sub 2}O{sub 7}. However, previous structural studies indicated that Tb{sub 2}Ti{sub 2}O{sub 7} is an imperfect pyrochlore. To clarify the situation, we performed neutron powder diffraction (NPD) and x-ray absorption fine structure (XAFS) measurements on samples that were prepared identically to those that show no magnetic order. The NPD measurements show that the long-range structure of Tb{sub 2}Ti{sub 2}O{sub 7} is well ordered with no structural transitions between 4.5 and 600 K. In particular, mean-squared displacements (u{sup 2}'s) for each site follow a Debye model with no offsets. No evidence for Tb/Ti site interchange was observed within an upper limit of 2%. Likewise, no excess or deficiency in the oxygen stoichiometry was observed, within an upper limit of 2% of the nominal pyrochlore value. Tb L{sub III} and Ti K-edge XAFS measurements from 20-300 K similarly indicate a well-ordered local structure. Other aspects of the structure are considered. We conclude that Tb{sub 2}Ti{sub 2}O{sub 7} has, within experimental error, an ideal, disorder-free pyrochlore lattice, thereby allowing the system to remain in a dynamic, frustrated spin state to the lowest observed temperatures.

  7. Microflow photochemistry: UVC-induced [2 + 2]-photoadditions to furanone in a microcapillary reactor

    Directory of Open Access Journals (Sweden)

    Sylvestre Bachollet

    2013-10-01

    Full Text Available [2 + 2]-Cycloadditions of cyclopentene and 2,3-dimethylbut-2-ene to furanone were investigated under continuous-flow conditions. Irradiations were conducted in a FEP-microcapillary module which was placed in a Rayonet chamber photoreactor equipped with low wattage UVC-lamps. Conversion rates and isolated yields were compared to analogue batch reactions in a quartz test tube. In all cases examined, the microcapillary reactor furnished faster conversions and improved product qualities.

  8. A simulation Model of the Reactor Hall Ventilation and air Conditioning Systems of ETRR-2

    International Nuclear Information System (INIS)

    Abd El-Rahman, M.F.

    2004-01-01

    Although the conceptual design for any system differs from one designer to another. each of them aims to achieve the function of the system required. the ventilation and air conditioning system of reactors hall is one of those systems that really differs but always dose its function for which it is designed. thus, ventilation and air conditioning in some reactor hall constitute only one system whereas in some other ones, they are separate systems. the Egypt Research Reactor-2 (ETRR-2)represents the second type. most studies conducted on ventilation and air conditioning simulation models either in traditional building or for research rectors show that those models were not designed similarly to the model of the hall of ETRR-2 in which ventilation and air conditioning constitute two separate systems.besides, those studies experimented on ventilation and air conditioning simulation models of reactor building predict the temperature and humidity inside these buildings at certain outside condition and it is difficult to predict when the outside conditions are changed . also those studies do not discuss the influences of reactor power changes. therefore, the present work deals with a computational study backed by infield experimental measurements of the performance of the ventilation and air conditioning systems of reactor hall during normal operation at different outside conditions as well as at different levels of reactor power

  9. Thermal design of heat-exchangeable reactors using a dry-sorbent CO2 capture multi-step process

    International Nuclear Information System (INIS)

    Moon, Hokyu; Yoo, Hoanju; Seo, Hwimin; Park, Yong-Ki; Cho, Hyung Hee

    2015-01-01

    The present study proposes a multi-stage CO 2 capture process that incorporates heat-exchangeable fluidized-bed reactors. For continuous multi-stage heat exchange, three dry regenerable sorbents: K 2 CO 3 , MgO, and CaO, were used to create a three-stage temperature-dependent reaction chain for CO 2 capture, corresponding to low (50–150 °C), middle (350–650 °C), and high (750–900 °C) temperature stages, respectively. Heat from carbonation in the high and middle temperature stages was used for regeneration for the middle and low temperature stages. The feasibility of this process is depending on the heat-transfer performance of the heat-exchangeable fluidized bed reactors as the focus of this study. The three-stage CO 2 capture process for a 60 Nm 3 /h CO 2 flow rate required a reactor area of 0.129 and 0.130 m 2 for heat exchange between the mid-temperature carbonation and low-temperature regeneration stages and between the high-temperature carbonation and mid-temperature regeneration stages, respectively. The reactor diameter was selected to provide dense fluidization conditions for each bed with respect to the desired flow rate. The flow characteristics and energy balance of the reactors were confirmed using computational fluid dynamics and thermodynamic analysis, respectively. - Highlights: • CO 2 capture process is proposed using a multi-stage process. • Reactor design is conducted considering heat exchangeable scheme. • Reactor surface is designed by heat transfer characteristics of fluidized bed

  10. 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%

  11. Core design calculations of IRIS reactor using modified CORD-2 code package

    International Nuclear Information System (INIS)

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

    2002-01-01

    Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)

  12. 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

  13. 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.

  14. 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

  15. BR2 reactor: medical and industrial applications

    International Nuclear Information System (INIS)

    Ponsard, B.

    2005-01-01

    The radioisotopes are produced for various applications in the nuclear medicine (diagnostic, therapy, palliation of metastatic bone pain), industry (radiography of welds, ...), agriculture (radiotracers, ...) and basic research. Due to the availability of high neutron fluxes (thermal neutron flux up to 10 15 n/cm 2 .s), the BR2 reactor is considered as a major facility through its contribution for a continuous supply of products such 99 Mo ( 99 mTc), 131 I, 133 Xe, 192 Ir, 186 Re, 153 Sm, 90 Y, 32 P, 188 W ( 188 Re), 203 Hg, 82 Br, 41 Ar, 125 I, 177 Lu, 89 Sr, 60 Co, 169 Yb, 147 Nd, and others. Neutron Transmutation Doped (NTD) silicon is produced for the semiconductor industry in the SIDONIE (Silicon Doping by Neutron Irradiation Experiment) facility, which is designed to continuously rotate and traverse the silicon through the neutron flux. These combined movements produce exceptional dopant homogeneity in batches of silicon measuring 4 and 5-inches in diameter by up to 750 mm in length. The main objectives of work performed were to provide a reliable and qualitative supply of radioisotopes and NTD-silicon to the customers in accordance with a quality system that has been certified to the requirements of the EN ISO 9001: 2000. This new Quality System Certificate has been obtained in November 2003 for the Production of radioisotopes for medical and industrial applications and the Production of Neutron Transmutation Doped (NTD) Silicon in the BR2 reactor

  16. Possible future roles for fast breeder reactors Part 1 and 2

    International Nuclear Information System (INIS)

    1978-06-01

    Part 1. The Fast Breeder Reactor (in particular in its sodium cooled version) has been steadily developed in the Community. This report attempts to quantify the advantages of this system in terms of fossil energy and uranium savings in the medium/long term as well as to examine some long term economic implications. The methodology of comparing scenarios, not individual reactor systems is followed. These scenarios have been chosen taking into account a range of assumptions concerning Community energy demand growth, fossil energy and uranium availability and technological capabilities. Part 2. The fast breeder reactor (FBR), particularly its sodium-cooled form (LMFBR) has been under development in the Community for many years. Industrial enterprises dedicated to its commercialisation have been formed and long range plans for its industrial utilisation are being formulated. The value of breeder reactors from the point of view of minimising Community fuel requirements has been discussed in Part I of this report (1). In Part II the consequences of delaying their introduction, and the demands placed upon the recycle industry by the introduction of fast reactors of different characteristics, using the Community electricity demand scenarios developed for Part I, are discussed. In addition comments are provided upon the effect of FBR introduction on the size of plutonium stocks

  17. A regression approach for Zircaloy-2 in-reactor creep constitutive equations

    International Nuclear Information System (INIS)

    Yung Liu, Y.; Bement, A.L.

    1977-01-01

    In this paper the methodology of multiple regressions as applied to Zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor Zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) When there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets. (2) Regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections. Multiple regression analysis performed on a set of carefully selected Zircaloy-2 in-reactor creep data leads to a model which provides excellent correlations for the data. (Auth.)

  18. Coolant radiolysis studies in the high temperature, fuelled U-2 loop in the NRU reactor

    International Nuclear Information System (INIS)

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

    2008-06-01

    An understanding of the radiolysis-induced chemistry in the coolant water of nuclear reactors is an important key to the understanding of materials integrity issues in reactor coolant systems. Significant materials and chemistry issues have emerged in Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR) and CANDU reactors that have required a detailed understanding of the radiation chemistry of the coolant. For each reactor type, specific computer radiolysis models have been developed to gain insight into radiolysis processes and to make chemistry control adjustments to address the particular issue. In this respect, modelling the radiolysis chemistry has been successful enough to allow progress to be made. This report contains a description of the water radiolysis tests performed in the U-2 loop, NRU reactor in 1995, which measured the CHC under different physical conditions of the loop such as temperature, reactor power and steam quality. (author)

  19. Design and computational analysis of passive siphon breaker for 49-2 swimming pool reactor

    International Nuclear Information System (INIS)

    Yue Zhiting; Song Yunpeng; Liu Xingmin; Zou Yao; Wu Yuanyuan

    2014-01-01

    Based on safety considerations, a passive siphon breaker will be added to the primary cooling system of 49-2 Swimming Pool Reactor (SPR). With the breaker location determined, the capability of siphon breakers with diameters of 1.5 cm and 2.0 cm was calculated and analyzed respectively by RELAP5/MOD3.3 code. The results show that in the condition of large break loss of coolant accident these two sizes of siphon breakers are able to break the siphon phenomena, and maintain the pool water level above the reactor core when the reactor and the pump are shutdown. In the end, to be conservative, the siphon breaker with diameter of 2.0 cm is adopted. (authors)

  20. Clinical Characteristics of Comorbid Narcissistic Personality Disorder in Patients With Borderline Personality Disorder.

    Science.gov (United States)

    Hörz-Sagstetter, Susanne; Diamond, Diana; Clarkin, John F; Levy, Kenneth N; Rentrop, Michael; Fischer-Kern, Melitta; Cain, Nicole M; Doering, Stephan

    2017-07-31

    This study examines psychopathology and clinical characteristics of patients with borderline personality disorder (BPD) and comorbid narcissistic personality disorder (NPD) from two international randomized controlled trials. From a combined sample of 188 patients with BPD, 25 also fulfilled criteria for a comorbid diagnosis of NPD according to DSM-IV. The BPD patients with comorbid NPD, compared to the BPD patients without comorbid NPD, showed significantly more BPD criteria (M = 7.44 vs. M = 6.55, p personality disorders, and were more likely to meet criteria for full histrionic PD diagnosis (44.0% vs. 14.2%, p disorders (M = 2.68 vs. M = 3.75, p = .033). No differences could be found in general functioning, self-harming behavior, and suicide attempts.

  1. Efficient H2O2/CH3COOH oxidative desulfurization/denitrification of liquid fuels in sonochemical flow-reactors.

    Science.gov (United States)

    Calcio Gaudino, Emanuela; Carnaroglio, Diego; Boffa, Luisa; Cravotto, Giancarlo; Moreira, Elizabeth M; Nunes, Matheus A G; Dressler, Valderi L; Flores, Erico M M

    2014-01-01

    The oxidative desulfurization/denitrification of liquid fuels has been widely investigated as an alternative or complement to common catalytic hydrorefining. In this process, all oxidation reactions occur in the heterogeneous phase (the oil and the polar phase containing the oxidant) and therefore the optimization of mass and heat transfer is of crucial importance to enhancing the oxidation rate. This goal can be achieved by performing the reaction in suitable ultrasound (US) reactors. In fact, flow and loop US reactors stand out above classic batch US reactors thanks to their greater efficiency and flexibility as well as lower energy consumption. This paper describes an efficient sonochemical oxidation with H2O2/CH3COOH at flow rates ranging from 60 to 800 ml/min of both a model compound, dibenzotiophene (DBT), and of a mild hydro-treated diesel feedstock. Four different commercially available US loop reactors (single and multi-probe) were tested, two of which were developed in the authors' laboratory. Full DBT oxidation and efficient diesel feedstock desulfurization/denitrification were observed after the separation of the polar oxidized S/N-containing compounds (S≤5 ppmw, N≤1 ppmw). Our studies confirm that high-throughput US applications benefit greatly from flow-reactors. Copyright © 2013 Elsevier B.V. All rights reserved.

  2. Direct In Situ Quantification of HO2 from a Flow Reactor.

    Science.gov (United States)

    Brumfield, Brian; Sun, Wenting; Ju, Yiguang; Wysocki, Gerard

    2013-03-21

    The first direct in situ measurements of hydroperoxyl radical (HO2) at atmospheric pressure from the exit of a laminar flow reactor have been carried out using mid-infrared Faraday rotation spectroscopy. HO2 was generated by oxidation of dimethyl ether, a potential renewable biofuel with a simple molecular structure but rich low-temperature oxidation chemistry. On the basis of the results of nonlinear fitting of the experimental data to a theoretical spectroscopic model, the technique offers an estimated sensitivity of reactor exit temperature range of 398-673 K. Accurate in situ measurement of this species will aid in quantitative modeling of low-temperature and high-pressure combustion kinetics.

  3. 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)

  4. Status and perspective of development of cold moderators at the IBR-2 reactor

    International Nuclear Information System (INIS)

    Kulikov, S; Shabalin, E

    2012-01-01

    The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams nos. 7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams nos. 2-3 and for beams nos. 1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (∼3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

  5. Status and perspective of development of cold moderators at the IBR-2 reactor

    Science.gov (United States)

    Kulikov, S.; Shabalin, E.

    2012-03-01

    The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams #7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams #2-3 and for beams #1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (~3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

  6. Comparing internal and alliance-based new product development processes: case studies in the food industry

    OpenAIRE

    Olsen, Nina Veflen; Gripsrud, Geir

    2011-01-01

    Companies may simultaneously pursue different new product development (NPD) strategies. This article reports a comparative two case design study of in-house NPD projects as well as alliance based NPD projects in a food company. Two contradicting proposition’s of the efficiency of NPD in an alliance compared to NPD performed internally are stated, and the findings indicate that the alliance based NPD solution creates a better context for NPD than the in-house solution. Less forwarding of unsol...

  7. 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.

  8. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  9. 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

  10. Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

    International Nuclear Information System (INIS)

    Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

    2003-01-01

    The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

  11. Further study on parameterization of reactor NAA: Pt. 2

    International Nuclear Information System (INIS)

    Tian Weizhi; Zhang Shuxin

    1989-01-01

    In the last paper, Ik 0 method was proposed for fission interference corrections. Another important kind of interferences in reator NAA is due to threshold reaction induced by reactor fast neutrons. In view of the increasing importance of this kind of interferences, and difficulties encountered in using the relative comparison method, a parameterized method has been introduced. Typical channels in heavy water reflector and No.2 horizontal channel of Heavy Water Research Reactor in the Insitute of Atomic Energy have been shown to have fast neutron energy distributions (E>4 MeV) close to primary fission neutron spectrum, by using multi-threshold detectors. On this basis, Ti foil is used as an 'instant fast neutron flux monitor' in parameterized corrections for threshold reaction interferences in the long irradiations. Constant values of φ f /φ s = 0.70 ± 0.02% have been obtained for No.2 rabbit channel. This value can be directly used for threshold reaction inference correction in the short irradiations

  12. EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

    Energy Technology Data Exchange (ETDEWEB)

    Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  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. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Memmott, Matthew; Boy, Guy; Charit, Indrajit; Manera, Annalisa; Downar, Thomas; Lee, John; Muldrow, Lycurgus; Upadhyaya, Belle; Hines, Wesley; Haghighat, Alierza

    2017-01-01

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project ''Integral Inherently Safe Light Water Reactors (I 2 S-LWR)''. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  15. Continuous backfitting measures for the FRG-1 and FRG-2 research reactors

    International Nuclear Information System (INIS)

    Blom, K.H.; Falck, K.; Krull, W.

    1990-01-01

    The GKSS-Research Centre Geesthacht GmbH has been operating the research reactors FRG-1 and FRG-2 with power levels of 5 MW and 15 MW for 31 and 26 years respectively. Safe operation at full power levels over so many years with an average utilization between 180 d to 250 d per year is possible only with great efforts in modernization and upgrading of the research reactors. Emphasis has been placed on backfitting since around 1975. At that time within the Federal Republic of Germany many new guidelines, rules, ordinances, and standards in the field of (power) reactor safety were published. Much work has been done on the modernization of the FRG-1 and FRG-2 research reactors therefore within the last ten years. Work done within the last two years and presently underway includes: measures against water leakage through the concrete and along the beam tubes; repair of both cooling towers; modernization of the ventilation system; measures for fire protection; activities in water chemistry and water quality; installation of a double tubing for parts of the primary piping of the FRG-1; replacement of instrumentation, process control systems (operation and monitoring system) and alarm system; renewal of the emergency power supply; installation of internal lightning protection; installation of a cold neutron source; enrichment reduction for FRG-1. These efforts will continue to allow safe operation of our research reactors over their whole operational life

  16. TiO2-photocatalyzed As(III) oxidation in a fixed-bed, flow-through reactor.

    Science.gov (United States)

    Ferguson, Megan A; Hering, Janet G

    2006-07-01

    Compliance with the U.S. drinking water standard for arsenic (As) of 10 microg L(-1) is required in January 2006. This will necessitate implementation of treatment technologies for As removal by thousands of water suppliers. Although a variety of such technologies is available, most require preoxidation of As(III) to As(V) for efficient performance. Previous batch studies with illuminated TiO2 slurries have demonstrated that TiO2-photocatalyzed AS(III) oxidation occurs rapidly. This study examined reaction efficiency in a flow-through, fixed-bed reactor that provides a better model for treatment in practice. Glass beads were coated with mixed P25/sol gel TiO2 and employed in an upflow reactor irradiated from above. The reactor residence time, influent As(III) concentration, number of TiO2 coatings on the beads, solution matrix, and light source were varied to characterize this reaction and determine its feasibility for water treatment. Repeated usage of the same beads in multiple experiments or extended use was found to affect effluent As(V) concentrations but not the steady-state effluent As(III) concentration, which suggests that As(III) oxidation at the TiO2 surface undergoes dynamic sorption equilibration. Catalyst poisoning was not observed either from As(V) or from competitively adsorbing anions, although the higher steady-state effluent As(III) concentrations in synthetic groundwater compared to 5 mM NaNO3 indicated that competitive sorbates in the matrix partially hinder the reaction. A reactive transport model with rate constants proportional to incident light at each bead layer fit the experimental data well despite simplifying assumptions. TiO2-photocatalyzed oxidation of As(III) was also effective under natural sunlight. Limitations to the efficiency of As(III) oxidation in the fixed-bed reactor were attributable to constraints of the reactor geometry, which could be overcome by improved design. The fixed-bed TiO2 reactor offers an environmentally

  17. Exxon nuclear neutronics design methods for pressurized water reactors. Supplement 2

    International Nuclear Information System (INIS)

    Skogen, F.B.; Stout, R.B.

    1977-01-01

    Modifications to the Exxon Nuclear PWR neutronic design calculational methods are presented as well as the results obtained when these improved methods are compared to reactor measurements. The basic PWR design tools remain unchanged; i.e., the XPOSE code is used for generating the basic nuclear parameters, the PDQ-7 code is used for calculating reactivity and x-y power distributions, and the XTG code is used for three-dimensional analysis. The recent start-up experiences at D. C. Cook Unit 1 and H. B. Robinson Unit 2 have provided a significant increase in the data base supporting the current ENC PWR neutronic methods. The verification comparisons contained in the supplement include reactor measurements from D. C. Cook Unit 1, Cycle 2; H. B. Robinson Unit 2, Cycles 4 and 5; Palisades Cycle 2, and R. E. Ginna, Cycle 7

  18. Characterization of fuel distribution in the Three Mile Island Unit 2 (TMI-2) reactor system by neutron and gamma-ray dosimetry

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McNeece, J.P.; Kaiser, B.J.; McElroy, W.N.

    1984-01-01

    Neutron and gamma-ray dosimetry are being used for nondestructive assessment of the fuel distribution throughout the Three Mile Island Unit 2 (TMI-2) reactor core region and primary cooling system. The fuel content of TMI-2 makeup and purification Demineralizer A has been quantified with Si(Li) continuous gamma-ray spectrometry and solid-state track recorder (SSTR) neutron dosimetry. For fuel distribution characterization in the core region, results from SSTR neutron dosimetry exposures in the TMI-2 reactor cavity are presented. These SSTR results are consistent with the presence of a significant amount of fuel debris, equivalent to several fuel assemblies or more, lying at the bottom of the reactor vessel. (Auth.)

  19. Estimation of power feedback parameters of the IBR-2M reactor by square wave reactivity

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Popov, A.K.; Sumkhuu, D.

    2016-01-01

    Parameters of the IBR-2M reactor power feedback (PFB) are estimated based on the analysis of power transients caused by deliberate square wave reactivity when the pulsed reactor operates in the self-regulation mode. The PFB of the IBR-2M is described by three linear first-order differential equations. Two components of the PFB are responsible for the negative feedback and one, for the positive. The overall feedback is negative, i.e., it has a stabilizing effect for the operation of the reactor. The slowest negative component of the PFB is probably caused by heating of the fuel. Periodically repeated in the process of exploitation, estimation of the PFB parameters is one of the methods to ensure safety operation of the reactor. [ru

  20. Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 3 of the Proceedings of 2. Yugoslav symposium on reactor physics includes three papers describing the following: model for spatial synthesis of automated control system of the GCR type reactor; model for analysis of hydrodynamic processes at the BHWR type reactors; mathematical model for safety analysis of heavy water power reactor

  1. Use of the TAT in the assessment of DSM-IV cluster B personality disorders.

    Science.gov (United States)

    Ackerman, S J; Clemence, A J; Weatherill, R; Hilsenroth, M J

    1999-12-01

    The Social Cognition and Object Relations Scale (SCORS), developed by Western, Lohr, Silk, Kerber, and Goodrich (1985), is a diagnostic instrument used to assess an array of psychological functioning by using clinical narratives such as the Thematic Apperception Test (TAT; Murray, 1943) stories. This study investigated the utility of the SCORS to differentiate between Diagnostic and Statistical Manual of Mental Disorders (4th ed. [DSM-IV]; American Psychiatric Association, 1994) antisocial personality disorder (ANPD), borderline personality disorder (BPD), narcissistic personality disorder (NPD), and Cluster C personality disorder (CPD). A sample of 58 patients was separated into four groups: ANPD (n = 9), BPD (n = 21; 18 with a primary BPD diagnosis and 3 with prominent borderline traits who met 4 of the 5 DSM-IV criteria necessary for a BPD diagnosis), NPD (n = 16; 8 with a primary NPD diagnosis and 8 with prominent narcissistic traits who met 4 of the 5 DSM-IV criteria necessary for a NPD diagnosis), and CPD (n = 12). These groups were then compared on the 8 SCORS variables by using 5 TAT cards (1, 2, 3BM, 4, and 13MF). Spearman-Brown correction for 2-way mixed effects model of reliability for the 8 SCORS variables ranged from .70 to .95. The results of categorical and dimensional analyses indicate that (a) SCORS variables can be used to differentiate ANPD, BPD, and NPD; (b) the BPD group scored significantly lower (greater maladjustment) than did the CPD group on certain variables; (c) the BPD group scored significantly lower (greater maladjustment) than did the NPD group on all 8 SCORS variables; (d) the ANPD group scored significantly lower than did the NPD group on certain variables; (e) certain variables were found to be empirically related to the total number of DSM-IV ANPD, BPD, and NPD criteria; and (f) certain variables were found to be empirically related to Minnesota Multiphasic Personality Inventory-2 (MMPI-2; Butcher, Dahlstrom, Graham, Tellegen

  2. TOKMINA, Toroidal Magnetic Field Minimization for Tokamak Fusion Reactor. TOKMINA-2, Total Power for Tokamak Fusion Reactor

    International Nuclear Information System (INIS)

    Hatch, A.J.

    1975-01-01

    1 - Description of problem or function: TOKMINA finds the minimum magnetic field, Bm, required at the toroidal coil of a Tokamak type fusion reactor when the input is beta(ratio of plasma pressure to magnetic pressure), q(Kruskal-Shafranov plasma stability factor), and y(ratio of plasma radius to vacuum wall radius: rp/rw) and arrays of PT (total thermal power from both d-t and tritium breeding reactions), Pw (wall loading or power flux) and TB (thickness of blanket), following the method of Golovin, et al. TOKMINA2 finds the total power, PT, of such a fusion reactor, given a specified magnetic field, Bm, at the toroidal coil. 2 - Method of solution: TOKMINA: the aspect ratio(a) is minimized, giving a minimum value for Bm. TOKMINA2: a search is made for PT; the value of PT which minimizes Bm to the required value within 50 Gauss is chosen. 3 - Restrictions on the complexity of the problem: Input arrays presently are dimensioned at 20. This restriction can be overcome by changing a dimension card

  3. System Definition Document: Reactor Data Necessary for Modeling Plutonium Disposition in Catawba Nuclear Station Units 1 and 2

    International Nuclear Information System (INIS)

    Ellis, R.J.

    2000-01-01

    The US Department of Energy (USDOE) has contracted with Duke Engineering and Services, Cogema, Inc., and Stone and Webster (DCS) to provide mixed-oxide (MOX) fuel fabrication and reactor irradiation services in support of USDOE's mission to dispose of surplus weapons-grade plutonium. The nuclear station units currently identified as mission reactors for this project are Catawba Units 1 and 2 and McGuire Units 1 and 2. This report is specific to Catawba Nuclear Station Units 1 and 2, but the details and materials for the McGuire reactors are very similar. The purpose of this document is to present a complete set of data about the reactor materials and components to be used in modeling the Catawba reactors to predict reactor physics parameters for the Catawba site. Except where noted, Duke Power Company or DCS documents are the sources of these data. These data are being used with the ORNL computer code models of the DCS Catawba (and McGuire) pressurized-water reactors

  4. Report on the operation in 1973 of the FR 2 research reactor

    International Nuclear Information System (INIS)

    Moeller, I.; Steiger, W.

    1975-04-01

    Also in 1973, the heavy-water moderated research and testing reactor FR 2 was operated to schedule at 44 MW nominal power. Again, the availability of the plant was slightly improved. Experimental utilization through instrumented irradiation capsules strongly increased as compared to the previous year. Up to 16 capsule test rigs at a time were inserted in the reactor. As to the beam tube experiments, up to 13 experiments covering a total of 18 test rigs were conducted simultaneously at the 12 reasonably usable beam holes. At the beginning of the year all of the positions available were occupied by 5 loop experiments. Isotope production reached its highest value with a total of 2,372 irradiated capsules (1.3% more than the year before). Some remarkable figures characterized the year 1973: On August 16, 1973 ten years of full power operation at a nominal power of 12 and 44 MW, respectively, had been reached. On July 24, 1973 the 50,000th isotope irradiation was performed in the reactor and on December 26, 1973 a total energy release of 100,000 MWd was recorded. Moreover, the 125,000th visitor of the reactor was welcomed on December 6, 1973. (orig./UA) [de

  5. Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor building decontamination. Summary status report. Volume 2

    International Nuclear Information System (INIS)

    Doerge, D.H.; Miller, R.L.; Scotti, K.S.

    1986-05-01

    This document summarizes information relating to decontamination of the Three Mile Island Unit 2 (TMI-2) reactor building. The report covers activities for the period of June 1, 1979 through March 29, 1985. The data collected from activity reports, reactor containment entry records, and other sources were entered into a computerized data system which permits extraction/manipulation of specific information which can be used in planning for recovery from an accident similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during decontamination of the reactor building. Support activities conducted outside of radiation areas are excluded from the scope of this report. Computerized reports included in this document are: a chronological summary listing work performed relating to reactor building decontamination for the period specified; and summary reports for each major task during the period. Each task summary is listed in chronological order for zone entry and subtotaled for the number of personnel entries, exposures, and man-hours. Manually-assembled table summaries are included for: labor and exposures by department and labor and exposures by major activity

  6. Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

    Energy Technology Data Exchange (ETDEWEB)

    Dougherty, D.; Adams, J. W.

    1983-08-01

    Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation.

  7. Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

    International Nuclear Information System (INIS)

    Dougherty, D.; Adams, J.W.

    1983-08-01

    Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation

  8. HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor

    International Nuclear Information System (INIS)

    Finch, D.R.

    1965-01-01

    1 - Description of problem or function: HERESY3 solves the two- dimensional, few-group, static reactor eigenvalue problem using the heterogeneous (source-sink or Feinburg-Galanin) formalism. The solution yields the reactor k-effective and absorption reaction rates for each rod normalized to the most absorptive rod in the thermal level. Epithermal fissions are allowed at each resonance level, and lattice-averaged values of thermal utilization, resonance escape probability, thermal and resonance eta values, and the fast fission factor are calculated. Kernels in the calculation are based on age-diffusion theory. Both finite reactor lattices and infinitely repeating reactor super-cells may be calculated. Rod parameters may be calculated by several internal options, and a direct interface is provided to a HAMMER system (NESC Abstract 277) lattice library tape to obtain cell parameters. Criticality searches are provided on thermal utilization, thermal eta, and axial leakage buckling. 2 - Method of solution: Direct power iteration on matrix form of the heterogeneous critical equation is used. 3 - Restrictions on the complexity of the problem: Maxima of - 50 flux/geometry symmetry positions; 20 physically different assemblies; 9 resonance levels; 5000 rod coordinate positions

  9. G2 and G3 reactors design; Description des reacteurs G2 et G3

    Energy Technology Data Exchange (ETDEWEB)

    Herreng,; Ertaud,; Pasquet, [Societe Alsacienne de Constructions Mecaniques (France)

    1958-07-01

    'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO{sub 2} under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO{sub 2} flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm{sup 2}). Steam can be condensed in the event of a group turbo-generator stopping, with no modifion for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO{sub 2}, its storage and drain. 49 boron carbide rods are used to control the

  10. Investigation of hydrogen-burn damage in the Three Mile Island Unit 2 reactor building

    International Nuclear Information System (INIS)

    Alvares, N.J.; Beason, D.G.; Eidem, G.R.

    1982-06-01

    About 10 hours after the March 28, 1979 Loss-of-Coolant Accident began at Three Mile Island Unit 2, a hydrogen deflagration of undetermined extent occurred inside the reactor building. Examinations of photographic evidence, available from the first fifteen entries into the reactor building, yielded preliminary data on the possible extent and range of hydrogen burn damage. These data, although sparse, contributed to development of a possible damage path and to an estimate of the extent of damage to susceptible reactor building items. Further information gathered from analysis of additional photographs and samples can provide the means for estimating hydrogen source and production rate data crucial to developing a complete understanding of the TMI-2 hydrogen deflagration. 34 figures

  11. A review of the probabilistic safety assessment application to the TR-2 research reactor

    International Nuclear Information System (INIS)

    Goektepe, G.; Adalioglu, U.; Anac, H.; Sevdik, B.; Menteseoglu, S.

    2001-01-01

    A review of the Probabilistic Safety Assessment (PSA) to the TR-2 Research Reactor is presented. The level 1 PSA application involved: selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development, dependent failure analysis. Each of the steps of the analysis given above is reviewed briefly with highlights from the selected results. PSA application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. Insights gained from the application of PSA methodology to the TR-2 research reactor led to a significant safety review of the system

  12. Solid-state track recorder neutron dosimetry in the Three-Mile Island Unit-2 reactor cavity

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.

    1985-04-01

    Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that there are at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity

  13. Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.; Rao, S.V.; Greenborg, J.; Fricke, V.R.

    1986-01-01

    Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity, for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. This existence of significant neutron streaming also explains the high count rate observed with the source range monitors that are located in the TMI-2 reactor cavity. (author)

  14. Degradation of gas-phase trichloroethylene over thin-film TiO2 photocatalyst in multi-modules reactor

    International Nuclear Information System (INIS)

    Kim, Sang Bum; Lee, Jun Yub; Kim, Gyung Soo; Hong, Sung Chang

    2009-01-01

    The present paper examined the photocatalytic degradation (PCD) of gas-phase trichloroethylene (TCE) over thin-film TiO 2 . A large-scale treatment of TCE was carried out using scale-up continuous flow photo-reactor in which nine reactors were arranged in parallel and series. The parallel or serial arrangement is a significant factor to determine the special arrangement of whole reactor module as well as to compact the multi-modules in a continuous flow reactor. The conversion of TCE according to the space time was nearly same for parallel and serial connection of the reactors.

  15. Top Management Involvement in New Product Development

    DEFF Research Database (Denmark)

    Felekoglu, Burcu; Maier, Anja; Moultrie, James

    2010-01-01

    a broader conceptual space than this participation. This paper reviews the literature on top management involvement in new product development (NPD) and discusses relevance of different theoretical perspectives from other disciplines such as management, organisational behaviour and communication to analyse......Involvement of top managers in new product development (NPD) is a critical factor affecting NPD performance and frequently considered to be the participation of top management to certain activities in NPD or their NPD related behaviours. However, “Top management involvement in NPD” occupies...... antecedents, realisation and consequences of top management involvement in NPD. It is argued that top management has different involvement at different NPD levels: organisation level and project level. Resulting from this literature review, a tentative framework for top management involvement in NPD...

  16. 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.

  17. 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.

  18. 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

  19. A gas-phase reactor powered by solar energy and ethanol for H2 production

    International Nuclear Information System (INIS)

    Ampelli, Claudio; Genovese, Chiara; Passalacqua, Rosalba; Perathoner, Siglinda; Centi, Gabriele

    2014-01-01

    In the view of H 2 as the future energy vector, we presented here the development of a homemade photo-reactor working in gas phase and easily interfacing with fuel cell devices, for H 2 production by ethanol dehydrogenation. The process generates acetaldehyde as the main co-product, which is more economically advantageous with respect to the low valuable CO 2 produced in the alternative pathway of ethanol photoreforming. The materials adopted as photocatalysts are based on TiO 2 substrates but properly modified with noble (Au) and not-noble (Cu) metals to enhance light harvesting in the visible region. The samples were characterized by BET surface area analysis, Transmission Electron Microscopy (TEM) and UV–visible Diffusive Reflectance Spectroscopy, and finally tested in our homemade photo-reactor by simulated solar irradiation. We discussed about the benefits of operating in gas phase with respect to a conventional slurry photo-reactor (minimization of scattering phenomena, no metal leaching, easy product recovery, etc.). Results showed that high H 2 productivity can be obtained in gas phase conditions, also irradiating titania photocatalysts doped with not-noble metals. - Highlights: • A gas-phase photoreactor for H 2 production by ethanol dehydrogenation was developed. • The photocatalytic behaviours of Au and Cu metal-doped TiO 2 thin layers are compared. • Benefits of operating in gas phase with respect to a slurry reactor are presented. • Gas phase conditions and use of not-noble metals are the best economic solution

  20. 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

  1. Comparing internal and alliance-based new product development processes: case studies in the food industry

    OpenAIRE

    Olsen, Nina Veflen; Gripsrud, Geir

    2011-01-01

    This is the authors’ final, accepted and refereed manuscript to the article Companies may simultaneously pursue different new product development (NPD) strategies. This article reports a comparative two case design study of in-house NPD projects as well as alliance based NPD projects in a food company. Two contradicting proposition’s of the efficiency of NPD in an alliance compared to NPD performed internally are stated, and the findings indicate that the alliance based NPD solution create...

  2. Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.; Rao, S.V.; Greenborg, J.; Fricke, V.R.

    1985-01-01

    Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity

  3. A Conceptual Study of a Supercritical CO2-Cooled Micro Modular Reactor

    Directory of Open Access Journals (Sweden)

    Hwanyeal Yu

    2015-12-01

    Full Text Available A neutronics conceptual study of a supercritical CO2-cooled micro modular reactor (MMR has been performed in this work. The suggested MMR is an extremely compact and truck-transportable nuclear reactor. The thermal power of the MMR is 36.2 MWth and it is designed to have a 20-year lifetime without refueling. A salient feature of the MMR is that all the components including the generator are integrated in a small reactor vessel. For a minimal volume and long lifetime of the MMR core, a fast neutron spectrum is utilized in this work. To enhance neutron economy and maximize the fuel volume fraction in the core, a high-density uranium mono-nitride U15N fuel is used in the fast-spectrum MMR. Unlike the conventional supercritical CO2-cooled fast reactors, a replaceable fixed absorber (RFA is introduced in a unique way to minimize the excess reactivity and the power peaking factor of the core. For a compact core design, the drum-type control absorber is adopted as the primary reactivity control mechanism. In this study, the neutronics analyses and depletions have been performed by using the continuous energy Monte Carlo Serpent code with the evaluated nuclear data file ENDF/B-VII.1 Library. The MMR core is characterized in view of several important safety parameters such as control system worth, fuel temperature coefficient (FTC and coolant void reactivity (CVR, etc. In addition, a preliminary thermal-hydraulic analysis has also been performed for the hottest channel of the Korea Advanced Institute of Science and Technology (KAIST MMR.

  4. Analysis of key hardware factors and countermeasure for restricting 49-2 swimming pool reactor lifetime

    International Nuclear Information System (INIS)

    Zhang Yadong; Guo Yue; Yang Xiao; Wang Yiwei; Wang Zhanwen

    2013-01-01

    Safe operation is the most important factor to determine the lifetime of aged 49-2 swimming pool reactor. In this paper, the hardware factors of lifetime were analyzed, such as the pool concrete aging, corrosion of aluminum container and primary coolant system, and graphite swelling etc., and then the corresponding measures such as surveillance, prevention and maintenance were purposed. The results show that 49-2 swimming pool reactor can continue to operate safely due to that container is safe under 8 degree earthquake, the reactor is safe on flood level of once per millennium, adding dam break, and the ageing condition of primary coolant system and container is acceptable. (authors)

  5. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    additional consideration should be required in nuclear design and fuel treating facilities due to reactivity coefficient being shifted to the plus side, larger neutron yield and increased heat source caused by MA loading. (2) Confirmation of TRU burning reactor core concepts. The core specification of sodium cooled-nitride fueled TRU burning large reactor was designed based on commercial type fast reactor (sodium cooled nitride fueled large fast reactor, 38000 MWt) which was designed in the feasibility studies on commercialized fast reactor cycle system. The composition of MAs from LWR's spent fuel was supposed. MA content in the core fuel is settled to 60 wt% based on the JAERI's design in order to maximize the MA transmutation amount. We need to exchange 25% of core fuel with zirconium hydride (ZrH 1.6 ) to attain Doppler coefficient being equivalent to that of the conventional type commercial fast reactor loaded 5 wt% MA. Furthermore, this reactor could transmute MAs produced in forty-eight sodium cooled nitride fueled large fast reactors generating the same output. In order to investigate the dependency of MA transmutation characteristics on the reactor output, 1200 MWt TRU burning middle or small reactor core concept was designed. This core was settled by reducing the number of core fuel assemblies from that of TRU burning large reactor designed above. MA transmutation rate in this core is smaller than that in the TRU burning large reactor core because the neutron flux of this core becomes smaller than that of the TRU burning large reactor core due to the higher Pu enrichment. (3) Comparison between TRU burning reactor and conventional type commercial fast reactor. MA transmutation and nuclear characteristics of the sodium cooled nitride fuel commercial type fast reactor loaded 5 wt%MA were evaluated and compared with those of TRU burning large reactor designed in (2). The commercial type fast reactor could only transmute MAs produced in seven sodium cooled nitride

  6. Empathy in narcissistic personality disorder: from clinical and empirical perspectives.

    Science.gov (United States)

    Baskin-Sommers, Arielle; Krusemark, Elizabeth; Ronningstam, Elsa

    2014-07-01

    Narcissistic personality disorder (NPD) is associated with an assortment of characteristics that undermine interpersonal functioning. A lack of empathy is often cited as the primary distinguishing feature of NPD. However, clinical presentations of NPD suggest that empathy is not simply deficient in these individuals, but dysfunctional and subject to a diverse set of motivational and situational factors. Consistent with this presentation, research illustrates that empathy is multidimensional, involving 2 distinct emotional and cognitive processes associated with a capacity to respectively understand and respond to others' mental and affective states. The goal of this practice review is to bridge the gap between our psychobiological understanding of empathy and its clinical manifestations in NPD. We present 3 case studies highlighting the variability in empathic functioning in people with NPD. Additionally, we summarize the literature on empathy and NPD, which largely associates this disorder with deficient emotional empathy, and dysfunctional rather than deficient cognitive empathy. Because this research is limited, we also present empathy-based findings for related syndromes (borderline and psychopathy). Given the complexity of narcissism and empathy, we propose that multiple relationships can exist between these constructs. Ultimately, by recognizing the multifaceted relationship between empathy and narcissism, and moving away from an all or nothing belief that those with NPD simply lack empathy, therapists may better understand narcissistic patients' behavior and motivational structure. (c) 2014 APA, all rights reserved.

  7. On-line reactor building integrity testing at Gentilly-2 (summary of results 1987-1994)

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1994-01-01

    In 1987, Hydro-0uebec embarked on an ambitious development program to provide the Gentilly-2 Nuclear Power Station with an effective and practical Reactor Building Containment integrity Test (CIT). In October 1992, the inaugural low pressure (3 kPa(g) nominal) CIT at 100% F.P was performed. The test was conclusive and the CIT was declared In-Service for containment integrity verification on-line. Five subsequent CITs performed in 1993 and 1994 have demonstrated the expected leak rate results and good reliability. The outstanding feature of the CITs is the demonstrated accurary of better than 5% of the measured leak rate. The CIT was developed with the primary goal of demonstrating 'overall' containment availability. Specifically it was designed to detect a 25 mm. diameter leak or hole in the Reactor Building. However, the remarkable CIT accuracy allows reliable detection of a 2 mm. hole. The Gentilly-2 CIT is an innovative approach based on the Temperature Compensation Method (TCM) which uses a reference volume composed of an extensive tubular network of several different diameters. This eliminates the need to track numerous temperature points. A second independent tubular network includes numerous humidity sampling points, thereby enabling the mearurernent of minute pressure variations inside the Reactor Building, independant of the spatial and temporal humidity behaviour. This Gentilly-2 TOM System has been demonstrated to work at both high and low test pressures. The GentiIly-2 design allows the CIT to be performed at a nominal 3 kPa(g) test pressure during a 12-hour period (28 hours total with alignment time) with the reactor at full power. The traditional Reactor Building Pressure Test (RBPT) is typically performed at high pressure (124 kPa(g) in a 5-day critical path window (7 days total with alignment time) during an annual shutdown

  8. Performance Estimation of Supercritical Co2 Micro Modular Reactor (MMR) for Varying Cooling Air Temperature

    International Nuclear Information System (INIS)

    Ahn, Yoonhan; Kim, Seong Gu; Cho, Seong Kuk; Lee, Jeong Ik

    2015-01-01

    A Small Modular Reactor (SMR) receives interests for the various application such as electricity co-generation, small-scale power generation, seawater desalination, district heating and propulsion. As a part of SMR development, supercritical CO2 Micro Modular Reactor (MMR) of 36.2MWth in power is under development by the KAIST research team. To enhance the mobility, the entire system including the power conversion system is designed for the full modularization. Based on the preliminary design, the thermal efficiency is 31.5% when CO2 is sufficiently cooled to the design temperature. A supercritical CO2 MMR is designed to supply electricity to the remote regions. The ambient temperature of the area can influence the compressor inlet temperature as the reactor is cooled with the atmospheric air. To estimate the S-CO2 cycle performance for various environmental conditions, A quasi-static analysis code is developed. For the off design performance of S-CO2 turbomachineries, the experimental result of Sandia National Lab (SNL) is utilized

  9. VENUS-2, Reactor Kinetics with Feedback, 2-D LMFBR Disassembly Excursions

    International Nuclear Information System (INIS)

    Jackson, J.F.; Nicholson, R.B.; Weber, D.P.

    1980-01-01

    1 - Description of problem or function: VENUS-2 is an improved edition of the VENUS fast-reactor disassembly program. It is a two- dimensional (r-z) coupled neutronics-hydrodynamics code that calculates the dynamic behavior of an LMFBR during a prompt-critical disassembly excursion. It calculates the power history and fission energy release as well as the space-time histories of the fuel temperatures, core material pressures, and core material motions. Reactivity feedback effects due to Doppler broadening and reactor material motion are taken into account. 2 - Method of solution: The power and energy release are calculated using a point-kinetics formulation with up to six delayed neutron groups. The reactivity is a combination of an input driving function and feedback effects due to Doppler broadening and material motion. An adiabatic model is used to calculate the temperature increase throughout the reactor based on an initial temperature distribution and power profile provided as input data. These temperatures are, in turn, converted to fuel pressures through one of several equation of state options provided. The material motion that results from the pressure buildup is calculated by a direct finite difference solution of a set of two-dimensional (r-z) hydrodynamics equations. This is done in Lagrangian coordinates. The reactivity change associated with this motion is calculated by first-order perturbation theory. The displacements are also used to adjust the fuel densities as required for the density dependent equation-of- state option. An automatic time-step-size selection scheme is provided. 3 - Restrictions on the complexity of the problem: VENUS-2 is written so that the dimensions of the storage arrays can be readily changed to accommodate a broad range of problem sizes. In the base version, the total number of mesh intervals is restricted such that (NR+3)*(NZ+3) is less than 700, where NR and NZ are the total number of mesh intervals in the r and z

  10. Nominal power density analysis of thermoelectric pins with non-constant cross sections

    International Nuclear Information System (INIS)

    Shi, Yaoguang; Mei, Deqing; Yao, Zhehe; Wang, Yancheng; Liu, Haiyan; Chen, Zichen

    2015-01-01

    Highlights: • Nominal power density of TEGs with non-constant cross sections pins is analyzed. • An analytical model of nominal power density (NPD) is developed. • Influences of shape parameter on NPD for different geometric pins are investigated. • Effects of dimensionless efficiency and the temperature ratio on NPD are examined. - Abstract: The investigation of the geometric structure of TEG (thermoelectric generator) pins is essential, as their geometry determines the performance of devices. In this study, nominal power density (NPD) is used to find a better geometric structure of thermoelectric pins of TEGs, since a comparison of maximum dimensionless efficiencies for different geometric pins cannot be used to identify the optimum geometry. The influence of shape parameter on NPD for TEG pins in linear, quadratic and exponential cross-sectional functions is studied. The NPD decreases when the shape parameter increases for different geometric pins, while the maximum values of NPD are the same. Then, the effects of dimensionless efficiency and the temperature ratio on the NPD are analyzed. The NPD decreases with the increase in dimensionless efficiency and temperature ratio. Pins with linear variation in cross section have the highest NPD among the three geometries of pins evaluated

  11. Energy Multiplier Module (EM{sup 2}) - advanced small modular reactor for electricity generation

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, T.; Schleicher, R.; Choi, H.; Rawls, J., E-mail: timothy.bertch@ga.com [General Atomics, San Diego, California (United States)

    2013-07-01

    In order to provide cost effective nuclear energy in other than large reactor, large grid applications, fission technology needs to make further advances. 'Convert and burn' fast reactors offer long life cores, improved fuel utilization, reduced waste and other benefits while achieving cost effective energy production in a smaller reactor. General Atomics' Energy Multiplier Module (EM{sup 2}), a helium-cooled compact fast reactor that augments its fissile fuel load with either depleted uranium (DU) or used nuclear fuel (UNF). The convert and burn in-situ provides 250 MWe with a 30 year core life. High temperature provides a simple, high efficiency direct cycle gas turbine which along with modular construction, fewer systems, road shipment and minimum on site construction support cost effectiveness. Additional advantages in fuel cycle, non-proliferation and siting flexibility and its ability to meet all safety requirements make for an attractive power source, especially in remote and small grid regions. (author)

  12. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [SCK CEN (Belgium); Kalcheva, S. [SCK CEN (Belgium); Sikik, E. [SCK CEN (Belgium); Koonen, E. [SCK CEN (Belgium)

    2015-12-01

    BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.

  13. Proceedings of the international topical meeting on advanced reactors safety: Volume 2

    International Nuclear Information System (INIS)

    1997-01-01

    In this volume, 89 papers are grouped under the following headings: advances in research/test reactor safety; advanced reactor accident management and emergency actions; advanced reactors instrumentation/controls/human factors; probabilistic risk/safety and reliability assessments; steam explosion research and issues; advanced reactor severe accident issues and research (analysis and assessments); advanced reactor thermal hydraulics; accelerator-driven source safety; liquid-metal reactor safety; structural assessments and issues; late papers

  14. 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)

  15. Application of 2DOF controller for reactor power control. Verification by numerical simulation

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Suzuki, Katsuo

    1996-09-01

    In this report the usefulness of the two degree of freedom (2DOF) control is discussed to improve the reference response characteristics and robustness for reactor power control system. The 2DOF controller consists of feedforward and feedback elements. The feedforward element was designed by model matching method and the feedback element by solving the mixed sensitivity problem of H ∞ control. The 2DOF control gives good performance in both reference response and robustness to disturbance and plant perturbation. The simulation of reactor power control was performed by digitizing the 2DOF controller with the digital control periods of 10[msec]. It is found that the control period of 10[msec] is enough not to make degradation of the control performance by digitizing. (author)

  16. Validation of SCALE4.4a for Calculation of Xe-Sm Transients After a Scram of the BR2 Reactor

    International Nuclear Information System (INIS)

    Kalcheva, S.; Ponsard, B.; Koonen, E.

    2007-01-01

    The aim of this report is to validate the computational modules system SCALE4.4a for evaluation of reactivity changes, macroscopic absorption cross sections and calculations of the positions of the Control Rods during their motion in Xe-Sm transient after a scram of the BR-2 reactor. The rapid shutting down of the reactor by inserting of negative reactivity by the Control Rods is known as a reactor scram. Following reactor scram, a large xenon and samarium buildup occur in the reactor, which may appreciably affect the multiplication factor of the core due to enormous neutron absorption. The validation of the calculations of Xe-Sm transients by SCALE4.4a has been performed on the measurements of the positions of the Control Rods during their motion in Xe-Sm transients of the BR-2 reactor and on comparison with the calculations by the standard procedure XESM, developed at the BR-2 reactor. A final conclusion is made that the SCALE4.4a modules system can be used for evaluation of Xe-Sm transients of the BR-2 reactor. The utilization of the code is simple, the computational time takes from few seconds.

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

    International Nuclear Information System (INIS)

    1981-06-01

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

  18. Degradation of gas-phase trichloroethylene over thin-film TiO{sub 2} photocatalyst in multi-modules reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Bum [New and Renewable Energy Team, Environment and Energy Division, Korea Institute of Industrial Technology (Korea, Republic of); Lee, Jun Yub, E-mail: ljy02191@hanafos.com [Power Engineering Research Institute, Korea Power Engineering Company, Inc. (Korea, Republic of); Kim, Gyung Soo [New and Renewable Energy Team, Environment and Energy Division, Korea Institute of Industrial Technology (Korea, Republic of); Hong, Sung Chang [Department of Environmental Engineering, Kyonggi University (Korea, Republic of)

    2009-07-30

    The present paper examined the photocatalytic degradation (PCD) of gas-phase trichloroethylene (TCE) over thin-film TiO{sub 2}. A large-scale treatment of TCE was carried out using scale-up continuous flow photo-reactor in which nine reactors were arranged in parallel and series. The parallel or serial arrangement is a significant factor to determine the special arrangement of whole reactor module as well as to compact the multi-modules in a continuous flow reactor. The conversion of TCE according to the space time was nearly same for parallel and serial connection of the reactors.

  19. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    Maillard, M.L.

    1960-01-01

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [fr

  20. Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

    2017-05-15

    Highlights: • Developed low enrichment and natural circulation cooled SLIMM-1.2 SMR for generating 10–100 MW{sub th}. • Neutronics analyses estimate operation life and temperature reactivity feedback. • At 100 MW{sub th}, SLIMM-1.2 operates for 6.3 FPY without refueling. • SLIMM-1.2 has relatively low power peaking and maximum UN fuel temperature < 1400 K. - Abstract: The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW{sub th} continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW{sub th}, the fissile production and depletion, the

  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. Development of the fast reactor group constant set JFS-3-J3.2R based on the JENDL-3.2

    CERN Document Server

    Chiba, G

    2002-01-01

    It is reported that the fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL3.2 has a serious error in the process of applying the weighting function. As the error affects greatly nuclear characteristics, and a corrected version of the reactor constant set, JFS-3-J3.2R, was developed, as well as lumped FP cross sections. The use of JFS-3-J3.2R improves the results of analyses especially on sample Doppler reactivity and reaction rate in the blanket region in comparison with those obtained using the JFS-3-J3.2.

  3. 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

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

    International Nuclear Information System (INIS)

    Potapov, I.A.; Serebrov, A.P.

    2001-01-01

    The reactor power, the required CNS dimensions and power of the cryogenic equipment define the CNS type with maximized cold neutron production. Cold neutron fluxes from liquid hydrogen (LH 2 ) and liquid deuterium (LD 2 ) cold neutron sources (CNS) are analyzed. Different CNS volumes, presents and absence of reentrant holes inside the CNS, different adjustment of beam tube and containment are considered. (orig.)

  5. Fusion reactor materials program plan. Section 2. Damage analysis and fundamental studies

    International Nuclear Information System (INIS)

    1978-07-01

    The scope of this program includes: (1) Development of procedures for characterizing neutron environments of test facilities and fusion reactors, (2) Theoretical and experimental investigations of the influence of irradiation environment on damage production, damage microstructure evolution, and mechanical and physical property changes, (3) Identification and, where appropriate, development of essential nuclear and materials data, and (4) Development of a methodology, based on damage mechanisms, for correlating the mechanical behavior of materials exposed to diverse test environments and projecting this behavior to magnetic fusion reactor (MFR) environments. Some major problem areas are addressed

  6. 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

  7. Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwangmin, E-mail: kwangmin81@gmail.com [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Go, Byeong-Soo; Sung, Hae-Jin; Park, Hea-chul; Kim, Seokho [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of); Jin, Yoon-Su; Oh, Yunsang [Vector Fields Korea Inc., Pohang 790-834 (Korea, Republic of); Park, Minwon [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Yu, In-Keun, E-mail: yuik@changwon.ac.kr [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of)

    2014-09-15

    Highlights: • The authors designed and fabricated a D-shape coil based toroid-type HTS DC reactor using 2G GdBCO HTS wires. • The toroid-type magnet consisted of 30 D-shape double pancake coil (DDC)s. The total length of the wire was 2.32 km. • The conduction cooling method was adopted for reactor magnet cooling. • The maximum cooling temperature of reactor magnet is 5.5 K. • The inductance was 408 mH in the steady-state condition (300 A operating). - Abstract: This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

  8. IAEA safety standards for research reactors

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

    The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

  9. Fusion pilot plant scoping study

    International Nuclear Information System (INIS)

    Gierszewski, P.J.; Blevins, P.J.; Brunnader, H.; Natalizio, A.; Cumyn, P.; Dean, B.; Smith, S.; Galambos, J.; Holloway, C.; Stremlaw, J.; Williams, G.

    1994-05-01

    CFFTP Pilot is representative of a class of machines that, like NPD in the CANDU development program, could test the key reactor core technologies on an integrated power reactor relevant system (materials, conditions, configuration). But in order to reduce costs, the machine would operate at reduced neutron flux relative to a power reactor, would not produce electricity, and would not test superconducting magnets. This design shows research directions towards a machine that could provide integrated nuclear testing (but not ignition physics) at a cost of about 1/3 ITER CDA. The test volume - the outboard blanket volume - would be comparable to the test port volume on ITER CDA, while the fluence and power density would be about 1/4 ITER CDA. 91 refs., 43 tabs., 45 figs

  10. Fusion pilot plant scoping study

    Energy Technology Data Exchange (ETDEWEB)

    Gierszewski, P J; Blevins, P J; Brunnader, H; Natalizio, A [Canadian Fusion Fuels Technology Project, Toronto, ON (Canada); Cumyn, P [Canatom Ltd., Montreal, PQ (Canada); Dean, B; Smith, S [Wardrop (W.L.) and Associates Ltd., Winnipeg, MB (Canada); Galambos, J [Oak Ridge National Lab., TN (United States); Holloway, C [Spar Aerospace Ltd., Toronto, ON (Canada); Stremlaw, J [Monenco AGRA Inc., Calgary, AB (Canada); Williams, G [Spectrum Engineering Corp., Peterborough, ON (Canada)

    1994-05-01

    CFFTP Pilot is representative of a class of machines that, like NPD in the CANDU development program, could test the key reactor core technologies on an integrated power reactor relevant system (materials, conditions, configuration). But in order to reduce costs, the machine would operate at reduced neutron flux relative to a power reactor, would not produce electricity, and would not test superconducting magnets. This design shows research directions towards a machine that could provide integrated nuclear testing (but not ignition physics) at a cost of about 1/3 ITER CDA. The test volume - the outboard blanket volume - would be comparable to the test port volume on ITER CDA, while the fluence and power density would be about 1/4 ITER CDA. 91 refs., 43 tabs., 45 figs.

  11. Containment Loads Analysis for CANDU6 Reactor using CONTAIN 2.0

    International Nuclear Information System (INIS)

    Kim, Tae H.; Yang, Chae Y.

    2013-01-01

    The containment plays an important role to limit the release of radioactive materials to the environment during design basis accidents (DBAs). Therefore, the containment has to maintain its integrity under DBA conditions. Generally, a containment functional DBA evaluation includes calculations of the key containment loads, i. e., pressure and temperature effects associated with a postulated large rupture of the primary or secondary coolant system piping. In this paper, the behavior of containment pressure and temperature was evaluated for loss of coolant accidents (LOCAs) of the Wolsong unit 1 in order to assess the applicability of CONTAIN 2.0 code for the containment loads analysis of the CANDU6 reactor. The containment pressure and temperature of the Wolsong unit 1 were evaluated using the CONTAIN 2.0 code and the results were compared with the CONTEMPT4 code. The peak pressure and temperature calculated by CONTAIN 2.0 agreed well with those of CONTEMPT4 calculation. The overall result of this analysis shows that the CONTAIN 2.0 code can apply to the containment loads analysis for the CANDU6 reactor

  12. Mass transfer of ammonia escape and CO2 absorption in CO2 capture using ammonia solution in bubbling reactor

    International Nuclear Information System (INIS)

    Ma, Shuangchen; Chen, Gongda; Zhu, Sijie; Han, Tingting; Yu, Weijing

    2016-01-01

    Highlights: • Mass transfer coefficient models of ammonia escape were built. • Influences of temperature, inlet CO 2 and ammonia concentration were studied. • Mass transfer coefficients of ammonia escape and CO 2 absorption were obtained. • Studies can provide the basic data as a reference guideline for process application. - Abstract: The mass transfer of CO 2 capture using ammonia solution in the bubbling reactor was studied; according to double film theory, the mass transfer coefficient models and interface area model were built. Through our experiments, the overall volumetric mass transfer coefficients were obtained, while the interface areas in unit volume were estimated. The volumetric mass transfer coefficients of ammonia escaping during the experiment were 1.39 × 10 −5 –4.34 × 10 −5 mol/(m 3 s Pa), and the volumetric mass transfer coefficients of CO 2 absorption were 2.86 × 10 −5 –17.9 × 10 −5 mol/(m 3 s Pa). The estimated interface area of unit volume in the bubbling reactor ranged from 75.19 to 256.41 m 2 /m 3 , making the bubbling reactor a viable choice to obtain higher mass transfer performance than the packed tower or spraying tower.

  13. Self- and informant-reported perspectives on symptoms of narcissistic personality disorder.

    Science.gov (United States)

    Cooper, Luke D; Balsis, Steve; Oltmanns, Thomas F

    2012-04-01

    Because narcissistic individuals tend to have an inflated view of themselves and their abilities, the reliance on self-reported information in the assessment and diagnosis of narcissistic personality disorder (NPD) is problematic. Hence, the use of informants in the assessment of NPD may be necessary. In the current study we examined self- and informant-reported features of NPD using agreement, frequency, and discrepancy analyses. The results indicated that informants tended to report more NPD features than selves, and that there were either low or nonsignificant levels of self-informant agreement among the 9 NPD diagnostic criteria and its categorical diagnosis. Informants were increasingly more likely to report higher raw scores relative to selves, indicating that the discrepancy between self- and informant reports increases with the NPD scale. Informants also reported NPD features that selves often did not, suggesting that current prevalence estimates of NPD, which use only self-reported information, are most likely underestimates. These results highlight the importance of gathering informant-reported data in addition to self-reported data when assessing NPD. (PsycINFO Database Record (c) 2013 APA, all rights reserved).

  14. SoLid: Search for Oscillations with Lithium-6 Detector at the SCK-CEN BR2 reactor

    Science.gov (United States)

    Ban, G.; Beaumont, W.; Buhour, J. M.; Coupé, B.; Cucoanes, A. S.; D'Hondt, J.; Durand, D.; Fallot, M.; Fresneau, S.; Giot, L.; Guillon, B.; Guilloux, G.; Janssen, X.; Kalcheva, S.; Koonen, E.; Labare, M.; Moortgat, C.; Pronost, G.; Raes, L.; Ryckbosch, D.; Ryder, N.; Shitov, Y.; Vacheret, A.; Van Mulders, P.; Van Remortel, N.; Weber, A.; Yermia, F.

    2016-04-01

    Sterile neutrinos have been considered as a possible explanation for the recent reactor and Gallium anomalies arising from reanalysis of reactor flux and calibration data of previous neutrino experiments. A way to test this hypothesis is to look for distortions of the anti-neutrino energy caused by oscillation from active to sterile neutrino at close stand-off (˜ 6- 8m) of a compact reactor core. Due to the low rate of anti-neutrino interactions the main challenge in such measurement is to control the high level of gamma rays and neutron background. The SoLid experiment is a proposal to search for active-to-sterile anti-neutrino oscillation at very short baseline of the SCK•CEN BR2 research reactor. This experiment uses a novel approach to detect anti-neutrino with a highly segmented detector based on Lithium-6. With the combination of high granularity, high neutron-gamma discrimination using 6LiF:ZnS(Ag) and precise localization of the Inverse Beta Decay products, a better experimental sensitivity can be achieved compared to other state-of-the-art technology. This compact system requires minimum passive shielding allowing for very close stand off to the reactor. The experimental set up of the SoLid experiment and the BR2 reactor will be presented. The new principle of neutrino detection and the detector design with expected performance will be described. The expected sensitivity to new oscillations of the SoLid detector as well as the first measurements made with the 8 kg prototype detector deployed at the BR2 reactor in 2013-2014 will be reported.

  15. A novel condensation reactor for efficient CO2 to methanol conversion for storage of renewable electric energy

    NARCIS (Netherlands)

    Bos, Martin Johan; Brilman, Derk Willem Frederik

    2015-01-01

    A novel reactor design for the conversion of CO2 and H2 to methanol is developed. The conversion limitations because of thermodynamic equilibrium are bypassed via in situ condensation of a water/methanol mixture. Two temperatures zones inside the reactor ensure optimal catalyst activity (high

  16. Quality assurance in the project of RECH-2 research reactor

    International Nuclear Information System (INIS)

    Goycolea Donoso, C.; Nino de Zepeda Schele, A.

    1989-01-01

    The implantation of a Quality Assurance Program for the design, supply, construction, installation, and testing of the RECH-2 research reactor, is described in this paper. The obtained results, demonstrate that a Quality Assurance Program constitutes a suitable mean to assure that the installation complies with the safety and reliability requirements. (author)

  17. When grandiosity and vulnerability collide: Implicit and explicit self-esteem in patients with narcissistic personality disorder.

    Science.gov (United States)

    Vater, Aline; Ritter, Kathrin; Schröder-Abé, Michela; Schütz, Astrid; Lammers, Claas-Hinrich; Bosson, Jennifer K; Roepke, Stefan

    2013-03-01

    Narcissistic personality disorder (NPD) is characterized by reports of grandiosity including exaggerated illusions of superiority and entitlement (DSM-IV-TR, APA, 2000). Based on clinical theories (e.g., Kernberg, 1975), many researchers argue that high explicit self-esteem in narcissists masks underlying implicit vulnerability (low implicit self-esteem). Conversely, based on social learning theories (i.e., Millon, 1981), people with NPD are characterized by implicit grandiosity (high implicit self-esteem). We test these competing hypotheses in patients diagnosed with NPD. The present study examined implicit self-esteem (using an Implicit Association Test) and explicit self-esteem (using a self-report questionnaire) in patients with NPD in comparison to non-clinical and clinical, non-NPD (Borderline Personality Disorder, BPD) control groups. Patients with NPD scored lower on explicit self-esteem than non-clinical controls. In comparison to patients with BPD, NPD patients scored higher on explicit and implicit self-esteem. Moreover, within the group of NPD patients, damaged self-esteem (i.e., low explicit, high implicit) was associated with higher narcissistic psychopathology. In both clinical groups we included participants seeking psychiatric treatment, which might influence explicit self-esteem. Longitudinal studies are needed to further assess self-esteem stability in NPD patients in comparison to the control groups. Our findings are indicative of vulnerable facets in patients with NPD (i.e., low explicit self-esteem). Furthermore, damaged self-esteem is connected to specific psychopathology within the NPD group. Implications for research on NPD are discussed. Copyright © 2012 Elsevier Ltd. All rights reserved.

  18. Set of rules SOR 2 licensing of nuclear reactors

    International Nuclear Information System (INIS)

    1976-05-01

    This is the set of rules promulgated by the Israel Atomic Energy Commission pursuant to the Supervision of Supplies and Services Law 5718-1957, Order regarding Supervision of Nuclear Reactors (1974) Chapter 3: Permits, to provide for the Licensing of Nuclear Reactors. (B.G.)

  19. Internationalization of New Product Development

    DEFF Research Database (Denmark)

    Brem, Alexander; Freitag, F.

    2014-01-01

    , reward systems, NPD creativity techniques, market orientation, and the average age of NPD teams. Other aspects, such as top management support, the use of structured NPD processes, and the use of heterogeneous NPD teams, show no substantial differences between the countries. Our findings suggest that...

  20. Measurement of thermal conductivity of sintered UO{sub 2} in the reactor; Merenje toplotne provodljivosti sinterovanog UO{sub 2} u reaktoru

    Energy Technology Data Exchange (ETDEWEB)

    Katanic, J; Stevanovic, M [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

    1965-10-15

    Thermal conductivity is considered one of the fundamental properties of sintered UO{sub 2} fuel. Samples should be tested under real core conditions. This paper covers the methods and instruments for thermal conductivity measurement of UO{sub 2} samples in the reactor core, measurements outside the core under conditions similar to those in the core and outside the core after irradiation. Fuel samples are placed in capsules for irradiation in the reactor in-core loops.

  1. Geomechanical Analysis of Underground Coal Gasification Reactor Cool Down for Subsequent CO2 Storage

    Science.gov (United States)

    Sarhosis, Vasilis; Yang, Dongmin; Kempka, Thomas; Sheng, Yong

    2013-04-01

    Underground coal gasification (UCG) is an efficient method for the conversion of conventionally unmineable coal resources into energy and feedstock. If the UCG process is combined with the subsequent storage of process CO2 in the former UCG reactors, a near-zero carbon emission energy source can be realised. This study aims to present the development of a computational model to simulate the cooling process of UCG reactors in abandonment to decrease the initial high temperature of more than 400 °C to a level where extensive CO2 volume expansion due to temperature changes can be significantly reduced during the time of CO2 injection. Furthermore, we predict the cool down temperature conditions with and without water flushing. A state of the art coupled thermal-mechanical model was developed using the finite element software ABAQUS to predict the cavity growth and the resulting surface subsidence. In addition, the multi-physics computational software COMSOL was employed to simulate the cavity cool down process which is of uttermost relevance for CO2 storage in the former UCG reactors. For that purpose, we simulated fluid flow, thermal conduction as well as thermal convection processes between fluid (water and CO2) and solid represented by coal and surrounding rocks. Material properties for rocks and coal were obtained from extant literature sources and geomechanical testings which were carried out on samples derived from a prospective demonstration site in Bulgaria. The analysis of results showed that the numerical models developed allowed for the determination of the UCG reactor growth, roof spalling, surface subsidence and heat propagation during the UCG process and the subsequent CO2 storage. It is anticipated that the results of this study can support optimisation of the preparation procedure for CO2 storage in former UCG reactors. The proposed scheme was discussed so far, but not validated by a coupled numerical analysis and if proved to be applicable it could

  2. Analysis of SBO accident and natural circulation of 49-2 swimming pool reactor

    International Nuclear Information System (INIS)

    Wu Yuanyuan; Liu Tiancai; Sun Wei

    2012-01-01

    The transient thermal hydraulic characteristics of 49-2 Swimming Pool Reactor (SPR) were analyzed by RELAP5/MOD3.3 code to verify the capability of natural circulation and minus reactivity feedback for accident mitigation under the condition of station blackout (SBO). Then, the effects on accident consequence and sequence for core channels and primary pumps were briefly discussed. The calculation results show that the reactor can be shutdown by the effect of minus reactivity feedback, and the residual heat can be removed through the stable natural circulation. Therefore, it demonstrates that the 49-2 SPR is safe during the accident of SBO. (authors)

  3. Status of IVO-FR2-Vg7 experiment for irradiation of fast reactor fuel rods

    International Nuclear Information System (INIS)

    Elbel, H.; Kummerer, K.; Bojarsky, K.; Lopez Jimenez, J.; Otero de la Gandara, J.L.

    1979-01-01

    Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid) concerning a Joint Irradiation Program of Fast Reactor Fuel Rods. The design of fuel rods in general is defined, and, in particular of those with a density 94% DT and diameter 7.6 mm up to a burn-up of 7% FIMA, to be irradiated in the FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used in this irradiation, other possibilities of irradiation in the reactor will also be described. (auth.)

  4. 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.)

  5. Revision of fast reactor group constant set JFS-3-J2

    International Nuclear Information System (INIS)

    Takano, Hideki; Kaneko, Kunio.

    1989-10-01

    To improve the fast reactor group constant set JFS-3-J2 to be applicable for high burnup reactor calculations, group constants for 155 fission product nuclides and the lumped group cross sections for four mother fission isotopes of U-235, U-238, Pu-239 and Pu-241 have been generated. Furthermore, the group constants for higher actinides such as Am and Cm have been produced on the basis of the JENDL-2 nuclear data, so as to be able to use for TRU-transmutation calculations. Benchmark test of this revised set has been performed by analysing the 21 fast critical experimental assemblies. Benchmark calculation system based on one-dimensional Sn-method has been developed to investigate the accuracy of one-dimensional diffusion calculations. Significant difference between the results obtained with the diffusion and transport calculations was observed for small cores and the assemblies with iron or nickel reflector. (author)

  6. UCLA program in reactor studies: The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

  7. Virtuality, communication, and new product team creativity : a social network perspective

    NARCIS (Netherlands)

    Leenders, RTAJ; van Engelen, JML; Kratzer, J

    2003-01-01

    Creativity is essential to the performance of new product development (NPD) teams. Since the creative NPD task requires teams to combine and integrate input from multiple NPD team members, the team's communication pattern is an important determinant of NPD team creativity. In the empirical part of

  8. N2O Catalytic Decomposition – from Laboratory Experiment to Industry Reactor

    Czech Academy of Sciences Publication Activity Database

    Obalová, L.; Jirátová, Květa; Karásková, K.; Chromčáková, Ž.

    2012-01-01

    Roč. 191, č. 1 (2012), s. 116-120 ISSN 0920-5861 R&D Projects: GA TA ČR TA01020336 Institutional support: RVO:67985858 Keywords : N2O * catalytic decomposition * fixed bed reactor Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 2.980, year: 2012

  9. 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

  10. Alteration of installation of reactors (alteration of No.1 and No.2 reactor facilities) in Oi Power Station, Kansai Electric Power Co., Inc

    International Nuclear Information System (INIS)

    1984-01-01

    The Nuclear Safety Commission reported to the Minister of International Trade and Industry on October 27, 1983, that the technical capability was recognized to be adequate, and the safety after the alteration of the installation of reactors was judged to be ensured. At the time of deliberation, the guidelines for examining the safety design and safety evaluation of LWR facilities for power generation were used. Regarding the change of the degree of enrichment of replacement fuel from 3.2 to 3.4 wt.%, the limiting conditions are satisfied in the replacement core, and the nuclear design is appropriate. Eight test fuel assemblies using UO 2 pellets containing gadolinia are charged in the core of No.2 reactor, and the irradiation of two cycles is carried out. As the result of the safety examination regarding this test, the propriety of the nuclear design and mechanical design of the test fuel assemblies was confirmed. This alteration does not exert influence on the result of safety analysis made so far. This report was decided by the Committee on Examination of Reactor Safety based on the conclusion of No.26 subcommittee. (Kako, I.)

  11. 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

    TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

  12. Plasma-catalyst hybrid reactor with CeO2/γ-Al2O3 for benzene decomposition with synergetic effect and nano particle by-product reduction.

    Science.gov (United States)

    Mao, Lingai; Chen, Zhizong; Wu, Xinyue; Tang, Xiujuan; Yao, Shuiliang; Zhang, Xuming; Jiang, Boqiong; Han, Jingyi; Wu, Zuliang; Lu, Hao; Nozaki, Tomohiro

    2018-04-05

    A dielectric barrier discharge (DBD) catalyst hybrid reactor with CeO 2 /γ-Al 2 O 3 catalyst balls was investigated for benzene decomposition at atmospheric pressure and 30 °C. At an energy density of 37-40 J/L, benzene decomposition was as high as 92.5% when using the hybrid reactor with 5.0wt%CeO 2 /γ-Al 2 O 3 ; while it was 10%-20% when using a normal DBD reactor without a catalyst. Benzene decomposition using the hybrid reactor was almost the same as that using an O 3 catalyst reactor with the same CeO 2 /γ-Al 2 O 3 catalyst, indicating that O 3 plays a key role in the benzene decomposition. Fourier transform infrared spectroscopy analysis showed that O 3 adsorption on CeO 2 /γ-Al 2 O 3 promotes the production of adsorbed O 2 - and O 2 2‒ , which contribute benzene decomposition over heterogeneous catalysts. Nano particles as by-products (phenol and 1,4-benzoquinone) from benzene decomposition can be significantly reduced using the CeO 2 /γ-Al 2 O 3 catalyst. H 2 O inhibits benzene decomposition; however, it improves CO 2 selectivity. The deactivated CeO 2 /γ-Al 2 O 3 catalyst can be regenerated by performing discharges at 100 °C and 192-204 J/L. The decomposition mechanism of benzene over CeO 2 /γ-Al 2 O 3 catalyst was proposed. Copyright © 2017 Elsevier B.V. All rights reserved.

  13. COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors

    International Nuclear Information System (INIS)

    Kaminaga, Masanori

    1994-03-01

    The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author)

  14. Ethanol production by immobilized yeast and its CO2 gas effects on a packed bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, G M; Choi, C Y; Choi, Y D; Han, M H

    1982-10-01

    Immobilised yeast trapped in an alginate matrix demonstrated maximum activity at 30 degrees C and showed no pH effect between 3 and 7. Substrate inhibition was observed at glucose concentrations above 8% but the immobilised cells retained 70% of their maximum activity at 20% glucose concentration. The operation stability of immobilised cells was lower in simple glucose solution than in the activation medium in which only 20% of the activity was lost after 10 days operation. Inactivated immobilised yeast beads were reactivated by incubation in activation medium without a significant increase in cell numbers in a bead. During the operation of the immobilised yeast in a packed bed reactor, CO/sub 2/ gas accumulation adversely affected the reactor performance. An ideal plus flow reactor, not taking into account the formation of CO/sub 2/ gas bubbles and the presence of mass trasnfer resistance, was simulated using a kinetic model for the production of ethanol and the simulation results were compared with the actual reactor performance to determine the CO/sub 2/ gas effect, quantitatively. Up to 45% of the substrate conversion was lost due to the accumulation of CO/sub 2/ gas bubbles in all cases. (Refs. 21).

  15. Power Quality Problems Mitigation using Dynamic Voltage Restorer in Egypt Thermal Research Reactor (ETRR-2)

    International Nuclear Information System (INIS)

    Kandil, T.; Ayad, N.M.; Abdel Haleam, A.; Mahmoud, M.

    2013-01-01

    Egypt thermal research reactor (ETRR-2) was subjected to several Power Quality Problems such as voltage sags/swells, harmonics distortion, and short interruption. ETRR-2 encompasses a wide range of loads which are very sensitive to voltage variations and this leads to several unplanned shutdowns of the reactor due to trigger of the Reactor Protection System (RPS). The Dynamic Voltage Restorer (DVR) has recently been introduced to protect sensitive loads from voltage sags and other voltage disturbances. It is considered as one of the most efficient and effective solution. Its appeal includes smaller size and fast dynamic response to the disturbance. This paper describes a proposal of a DVR to improve power quality in ETRR-2 electrical distribution systems . The control of the compensation voltage is based on d-q-o algorithm. Simulation is carried out by Matlab/Simulink to verify the performance of the proposed method

  16. Modernization of turbine control system and reactor control system in Almaraz 1 and 2; MOdernizacion de los sistemas de control de turbina y del reactor en Almaraz 1 y 2

    Energy Technology Data Exchange (ETDEWEB)

    Pulido, C.; Diez, J.; Carrasco, J. A.; Lopez, L.

    2005-07-01

    The replacement of the Turbine Control System and Reactor Control System are part of the Almaraz modernization program for the Instrumentation and Control. For these upgrades Almaraz has selected the Ovation Platform that provides open architecture and easy expansion to other systems, these platforms is highly used in many nuclear and thermal plants around the world. One of the main objective for this project were to minimize the impact on the installation and operation of the plant, for that reason the project is implemented in two phases, Turbine Control upgrade and Reactor Control upgrade. Another important objective was to increase the reliability of the control system making them fully fault tolerant to single failures. The turbine Control System has been installed in Units 1 and 2 while the Reactor Control System will be installed in 2006 and 2007 outages. (Author)

  17. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  18. Managing innovation in SMEs product development in small Irish electronics firms

    CERN Document Server

    Ledwith, Ann; Bessant, John

    2009-01-01

    Innovating and developing new products is critical for the survival and growth of any small firm but particularly for technology based firms. This study of NPD at small Irish electronics firms makes two main contributions to knowledge. Firstly, the management of NPD at small firms is found to be different from that at large firms in several respects including; NPD resources and expenditure, organising for NPD, NPD process proficiency, marketing and technical skills and proficiency, R&D/marke...

  19. Study of the obtainment of Mo_2C by gas-solid reaction in a fixed and rotary bed reactor

    International Nuclear Information System (INIS)

    Araujo, C.P.B. de; Souza, C.P. de; Souto, M.V.M.; Barbosa, C.M.; Frota, A.V.V.M.

    2016-01-01

    Carbides' synthesis via gas-solid reaction overcomes many of the difficulties found in other processes, requiring lower temperatures and reaction times than traditional metallurgic routes, for example. In carbides' synthesis in fixed bed reactors (FB) the solid precursor is permeated by the reducing/carburizing gas stream forming a packed bed without mobility. The use of a rotary kiln reactor (RK) adds a mixing character to this process, changing its fluid-particle dynamics. In this work ammonium molybdate was subjected to carbo-reduction reaction (CH4 / H2) in both reactors under the same gas flow (15L / h) and temperature (660 ° C) for 180 minutes. Complete conversion was observed Mo2C (dp = 18.9nm modal particles sizes' distribution) in the fixed bed reactor. In the RK reactor this conversion was only partial (∼ 40%) and Mo2C and MoO3 (34nm dp = bimodal) could be observed on the produced XRD pattern. Partial conversion was attributed to the need to use higher solids loading in the reactor CR (50% higher) to avoid solids to centrifuge. (author)

  20. 2-DB, 2-D Multigroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search

    International Nuclear Information System (INIS)

    Little, W.W. Jr.; Hardie, R.W.; Hirons, T.J.; O'Dell, R.D.

    1969-01-01

    1 - Description of problem or function: 2DB is a flexible, two- dimensional (x-y, r-z, r-theta, hex geometry) diffusion code for use in fast reactor analyses. The code can be used to: (a) Compute fuel burnup using a flexible material shuffling scheme. (b) Perform criticality searches on time absorption (alpha), material concentrations, and region dimensions using a regular or adjoint model. Criticality searches can be performed during burnup to compensate for fuel depletion. (c) Compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Standard source-iteration techniques are used. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy (group) averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes, are formed by the user. The code does not contain built-in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated. The current 1108 version, however, is nominally restricted to 50 energy groups in a 65 K memory. In the 6600 version the power fraction, average burnup rate, and breeding ratio calculations are limited to reactors with a maximum of 50 zones

  1. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Memmott, Matthew [Brigham Young Univ., Provo, UT (United States); Boy, Guy [Florida Inst. of Technology, Melbourne, FL (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Lee, John [Univ. of Michigan, Ann Arbor, MI (United States); Muldrow, Lycurgus [Morehouse College, Atlanta, GA (United States); Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, Wesley [Univ. of Tennessee, Knoxville, TN (United States); Haghighat, Alierza [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States)

    2017-10-02

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I2S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  2. CANDU: study and review

    International Nuclear Information System (INIS)

    Morad, César M.; Santos, Thiago A. dos

    2017-01-01

    The CANDU (Canadian Deuterium Uranium) is a nuclear reactor developed by AECL (Atomic Energy of Canada Limited). The first small-scale reactor is known as NPD and was made in 1955 and commenced operation in 1962. It is a pressurized heavy water reactor and uses D2O as moderator and coolant and therefore uses natural uranium as fuel. There have been two major types of CANDU reactors, the original design of around 500 MWe that was intended to be used in multi-reactor installations in large plants, and the rationalized CANDU6 which has units in Argentina, South Korea, Pakistan, Romania and China. Throughout the 1980s and 90s the nuclear power market suffered a major crash, with few new plants being constructed in North America or Europe. Design work continued through, however, and a number of new design concepts were introduced that dramatically improved safety, capital costs, economics and overall performance. These Generation III+ and Generation IV machines became a topic of considerable interest in the early 2000s as it appeared a nuclear renaissance was underway and large numbers of new reactors would be built over the next decade. The present work aims to study the reactors of the CANDU type, exploring from its creation to studies directed to G-III and G-IV reactors. (author)

  3. CANDU: study and review

    Energy Technology Data Exchange (ETDEWEB)

    Morad, César M., E-mail: cesar.morad@usp.br [Universidade de São Paulo (POLI/USP), SP (Brazil). Escola Politécnica; Stefani, Giovanni L. de, E-mail: giovanni.stefani@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Santos, Thiago A. dos, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil)

    2017-07-01

    The CANDU (Canadian Deuterium Uranium) is a nuclear reactor developed by AECL (Atomic Energy of Canada Limited). The first small-scale reactor is known as NPD and was made in 1955 and commenced operation in 1962. It is a pressurized heavy water reactor and uses D2O as moderator and coolant and therefore uses natural uranium as fuel. There have been two major types of CANDU reactors, the original design of around 500 MWe that was intended to be used in multi-reactor installations in large plants, and the rationalized CANDU6 which has units in Argentina, South Korea, Pakistan, Romania and China. Throughout the 1980s and 90s the nuclear power market suffered a major crash, with few new plants being constructed in North America or Europe. Design work continued through, however, and a number of new design concepts were introduced that dramatically improved safety, capital costs, economics and overall performance. These Generation III+ and Generation IV machines became a topic of considerable interest in the early 2000s as it appeared a nuclear renaissance was underway and large numbers of new reactors would be built over the next decade. The present work aims to study the reactors of the CANDU type, exploring from its creation to studies directed to G-III and G-IV reactors. (author)

  4. Operating reactors licensing actions summary. Vol. 4, No. 2

    International Nuclear Information System (INIS)

    1984-04-01

    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

  5. Experimental estimations of the kinetics parameters of the IBR-2M reactor by stochastic noises

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Tajybov, L.A.; Garibov, A.A.; Mekhtieva, R.N.

    2012-01-01

    Experimental investigations of stochastic fluctuations of pulse energy of the IBR-2M reactor have been carried out which allowed us to obtain some of the parameters of the reactor kinetics. At different levels of average power a sequence of values of pulse energy was recorded with the calculation of the distribution parameters. An ionization chamber with boron installed near the active zone was used as a neutron detector. The research results allowed us to estimate the average lifetime of prompt neutrons τ = (6.53±0.2)·10 -8 s, absolute power of the reactor and intensity of the source of spontaneous neutrons S sp ≤(6.72±0.12)·10 6 s -1 . It was shown that the experimental results are close to the calculated ones

  6. Enhancements to the SLOWPOKE-2 nuclear research reactor at the Royal Military College of Canada

    Energy Technology Data Exchange (ETDEWEB)

    Hungler, P.C.; Andrews, M.T.; Weir, R.D.; Nielson, K.S.; Chan, P.K.; Bennett, L.G.I., E-mail: paul.hungler@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

    2014-07-01

    In 1985 a Safe Low Power C(K)ritical Experiment (SLOWPOKE) nuclear research reactor was installed at the Royal Military College of Canada (RMCC). The reactor at nominally 20 kW thermal was named SLOWPOKE-2 and the core was designed to have a total of 198 fuel pins with Low Enriched Uranium (LEU) fuel (19.89% U-235). Installation of the reactor was intended to provide an education tool for members of the Canadian Armed Forces (CAF) and an affordable neutron source for the application of neutron activation analysis (NAA) and radioisotope production. Today, the SLOWPOKE-2 at RMCC continues to be a key education tool for undergraduate and post-graduate students and successfully conducts NAA and isotope production as per its original design intent. RMCC has significantly upgraded the facility and instruments to develop capabilities such as delayed neutron and gamma counting (DNGC) and neutron imaging, including 2D thermal neutron radiography and 3D thermal neutron tomography. These unique nuclear capabilities have been applied to relevant issues in the CAF. The analog control system originally installed in 1985 has been removed and replaced in 2001 by the SLOWPOKE Integrated Reactor Control and Instrumentation System (SIRCIS) which is a digital controller. This control system continues to evolve with SIRCIS V2 currently in operation. The continual enhancement of the facility, instruments and systems at the SLOWPOKE-2 at RMCC will be discussed, including an update on RMCC's refueling plan. (author)

  7. Enhancements to the SLOWPOKE-2 nuclear research reactor at the Royal Military College of Canada

    International Nuclear Information System (INIS)

    Hungler, P.C.; Andrews, M.T.; Weir, R.D.; Nielson, K.S.; Chan, P.K.; Bennett, L.G.I.

    2014-01-01

    In 1985 a Safe Low Power C(K)ritical Experiment (SLOWPOKE) nuclear research reactor was installed at the Royal Military College of Canada (RMCC). The reactor at nominally 20 kW thermal was named SLOWPOKE-2 and the core was designed to have a total of 198 fuel pins with Low Enriched Uranium (LEU) fuel (19.89% U-235). Installation of the reactor was intended to provide an education tool for members of the Canadian Armed Forces (CAF) and an affordable neutron source for the application of neutron activation analysis (NAA) and radioisotope production. Today, the SLOWPOKE-2 at RMCC continues to be a key education tool for undergraduate and post-graduate students and successfully conducts NAA and isotope production as per its original design intent. RMCC has significantly upgraded the facility and instruments to develop capabilities such as delayed neutron and gamma counting (DNGC) and neutron imaging, including 2D thermal neutron radiography and 3D thermal neutron tomography. These unique nuclear capabilities have been applied to relevant issues in the CAF. The analog control system originally installed in 1985 has been removed and replaced in 2001 by the SLOWPOKE Integrated Reactor Control and Instrumentation System (SIRCIS) which is a digital controller. This control system continues to evolve with SIRCIS V2 currently in operation. The continual enhancement of the facility, instruments and systems at the SLOWPOKE-2 at RMCC will be discussed, including an update on RMCC's refueling plan. (author)

  8. Collective occupational dose for nuclear reactors of the 2., 3. and 4. generation

    International Nuclear Information System (INIS)

    Guidez, J.; Saturnin, A.

    2016-01-01

    In France during reactor operation the individual occupational doses are collected and recorded according to the law. When you sum up all the individual doses you get the yearly collective dose expressed in Man.Sv/year. This piece of information can be used to make comparisons between various types of reactors and between reactors of the same type. The results show a steady decrease of the collective dose for all types of reactors over the time except for CANDU reactors for which a slight increase of the dose has appeared since the years 1996-1998. The decrease is due to the continuous improvement of reactor operating and to changes in the reactor design. There is also a constant gap over time between the collective dose for a BWR reactor (1.12 Man.Sv/y) and a PWR reactor 0.60 Man.Sv/y), this gap is certainly due to N 16 nuclide that is created in the primary circuit and transported to turbines in the case of a BWR reactor. For sodium-cooled fast reactors (RNR-Na) the collective dose is below 0.40 Man.Sv/y except for the BN-600 reactor. (A.C.)

  9. 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

  10. Acid sphingomyelinase (Asm) deficiency patients in The Netherlands and Belgium: disease spectrum and natural course in attenuated patients.

    Science.gov (United States)

    Hollak, C E M; de Sonnaville, E S V; Cassiman, D; Linthorst, G E; Groener, J E; Morava, E; Wevers, R A; Mannens, M; Aerts, J M F G; Meersseman, W; Akkerman, E; Niezen-Koning, K E; Mulder, M F; Visser, G; Wijburg, F A; Lefeber, D; Poorthuis, B J H M

    2012-11-01

    Niemann-Pick disease (NPD) is a neurovisceral lysosomal storage disorder caused by acid sphingomyelinase (ASM) deficiency, which can be categorized as either Niemann-Pick disease type A [NPD-A], with progressive neurological disease and death in early childhood, or as Niemann-Pick disease type B [NPD-B], with a more variable spectrum of manifestations. Enzyme replacement therapy (ERT) with recombinant sphingomyelinase is currently studied as potential treatment for NPD-B patients. The objective of this study is to characterize the clinical features of patients with ASM deficiency in the Netherlands and Belgium with focus on the natural disease course of NPD-B patients. Prospective and retrospective data on ASM deficient patients were collected in The Netherlands and part of Belgium. Patients with NPD-B that could be followed prospectively were evaluated every 6-12 months for pulmonary function tests, 6 minute walk test (6 MWT), imaging (bone marrow infiltration measured by QCSI, organ volumes by MRI and CT scan of the lungs) and biochemical markers. Twenty-five patients with ASM deficiency were identified (13 males, 12 females, median age 13years, range 1-59 years). Nine patients had died at the time of the study, including four NPD-A patients at the age of 1,1, 2, 3 and five NPDB patents at the age of 5, 6, 43, 56 and 60 years. There was a high prevalence of homozygosity and compound heterozygosity for the common p.Arg608del mutation in 43% and 19% of NPD-B patients, respectively. In NPD-B patients, thrombocytopenia was present in most, while anemia and leucopenia were less common (33% and 6 % respectively). HDL cholesterol was reduced in most patients. Pulmonary disease was severe in several patients. Follow-up up to 11 years revealed a gradual decrease in platelet count. Detailed investigations in 6 NPD-B patients with follow-up in 4 patients revealed remarkable stable disease parameters up to 6 years, with some decline in pulmonary function and 6 MWT. Bone

  11. Narcissistic personality disorder: relations with distress and functional impairment.

    Science.gov (United States)

    Miller, Joshua D; Campbell, W Keith; Pilkonis, Paul A

    2007-01-01

    This study examined the construct validity of narcissistic personality disorder (NPD) by examining the relations between NPD and measures of psychologic distress and functional impairment both concurrently and prospectively across 2 samples. In particular, the goal was to address whether NPD typically "meets" criterion C of the DSM-IV definition of Personality Disorder, which requires that the symptoms lead to clinically significant distress or impairment in functioning. Sample 1 (n = 152) was composed of individuals receiving psychiatric treatment, whereas sample 2 (n = 151) was composed of both psychiatric patients (46%) and individuals from the community. Narcissistic personality disorder was linked to ratings of depression, anxiety, and several measures of impairment both concurrently and at 6-month follow-up. However, the relations between NPD and psychologic distress were (a) small, especially in concurrent measurements, and (b) largely mediated by impaired functioning. Narcissistic personality disorder was most strongly related to causing pain and suffering to others, and this relationship was significant even when other Cluster B personality disorders were controlled. These findings suggest that NPD is a maladaptive personality style which primarily causes dysfunction and distress in interpersonal domains. The behavior of narcissistic individuals ultimately leads to problems and distress for the narcissistic individuals and for those with whom they interact.

  12. 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

  13. Safety of nuclear power reactors

    International Nuclear Information System (INIS)

    MacPherson, H.G.

    1982-01-01

    Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

  14. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-07-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities.

  15. 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

  16. 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

  17. Physical measurements at the RA reactor related to VISA-2, e. Measurements of flux and reactivity during RA reactor operation and exploitation; Fizicka merenja na reaktoru RA u vezi projekta VISA-2, e. Pracenje fluksa i reaktivnosti u toku eksploatacije reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-05-15

    This report includes the following: characteristics of neutron flux in vertical experimental channels of the RA reactor; characteristics of neutron flux in VISA-2 channels; reactivity changes in the reactor during VISA-2 irradiation including calibration of control rods.

  18. Passive Decay Heat Removal System Options for S-CO2 Cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Moon, Jangsik; Jeong, Yong Hoon; Lee, Jeong Ik

    2014-01-01

    To achieve modularization of whole reactor system, Micro Modular Reactor (MMR) which has been being developed in KAIST took S-CO 2 Brayton power cycle. The S-CO 2 power cycle is suitable for SMR due to high cycle efficiency, simple layout, small turbine and small heat exchanger. These characteristics of S-CO 2 power cycle enable modular reactor system and make reduced system size. The reduced size and modular system motived MMR to have mobility by large trailer. Due to minimized on-site construction by modular system, MMR can be deployed in any electricity demand, even in isolated area. To achieve the objective, fully passive safety systems of MMR were designed to have high reliability when any offsite power is unavailable. In this research, the basic concept about MMR and Passive Decay Heat Removal (PDHR) system options for MMR are presented. LOCA, LOFA, LOHS and SBO are considered as DBAs of MMR. To cope with the DBAs, passive decay heat removal system is designed. Water cooled PDHR system shows simple layout, but has CCF with reactor systems and cannot cover all DBAs. On the other hand, air cooled PDHR system with two-phase closed thermosyphon shows high reliability due to minimized CCF and is able to cope with all DBAs. Therefore, the PDHR system of MMR will follows the air-cooled PDHR system and the air cooled system will be explored

  19. A study of UO2 wafer fuel for very high-power research reactors

    International Nuclear Information System (INIS)

    Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.

    1983-01-01

    The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to 2 caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO 2 wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full power operation up to a burnup, of 1.9x10 21 fis/cm 3 ; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, the wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. (author)

  20. Microstructure in Zircaloy Creep Tested in the R2 Reactor

    International Nuclear Information System (INIS)

    Pettersson, Kjell

    2004-12-01

    Tubular specimens of Zircaloy-4 have been creep tested in bending in the R2 reactor in Studsvik. The creep deformation in the reactor core is accelerated in comparison with creep deformation outside the reactor core. The possible mechanisms behind this behaviour are described briefly. In order to determine which the actual mechanism is, the microstructure of the material creep tested in the R2 reactor has been examined by transmission electron microscopy. Due to the bending, material subjected to both tensile and compressive stress during creep was available. Since some of the proposed mechanisms might give microstructures which are different when the material is subjected to compressive or tensile stress it was assumed that examination of both types of material would give valuable information with regard to the operating mechanism. The result of the examination was that in the as-irradiated condition there were no obvious differences detected between materials which had been deformed in tension or compression. After a heat treatment to coarsen the irradiation induced microstructure there were still no significant differences between the two types of material. However it was now observed that in addition to dislocation loops the microstructure also contained network dislocations which presumably had been invisible in the electron microscope before heat treatment due to the high density of small dislocation loops in this state. It is therefore concluded that the most probable mechanism for irradiation creep in this case is climb and glide of the network dislocations. The role of irradiation is two-fold: It accelerates climb due to the production of point defects of which more interstitials than vacancies arrive to the network dislocations stopped at an obstacles. This leads to a net climb after which a dislocation is released from the obstacle and an amount of glide takes place. The second effect is the production of loops which serve as an increasing density of

  1. In-situ stripping of H{sub 2}S in gasoil hydrodesulphurization - reactor design considerations

    Energy Technology Data Exchange (ETDEWEB)

    Nava, J.A.O.; Krishna, R. [Amsterdam Univ., Dept. of Chemical Engineering, Amsterdam (Netherlands)

    2004-02-01

    In order to meet future diesel specifications the sulphur content of diesel would need to be reduced to below 50 ppm. This requirement would require improved reactor configurations. In this study we examine the benefits of counter-current contacting of gas oil with H{sub 2}, over conventional co-current contacting in a trickle bed hydrodesulphurization (HDS) reactor. In counter-current contacting, we achieve in-situ stripping of H{sub 2}S from the liquid phase; this is beneficial to the HDS kinetics. A comparison simulation study shows that counter-current contacting would require about 20% lower catalyst load than co-current contacting. However, counter-current contacting of gas and liquid phases in conventionally used HDS catalysts, of 1.5 mm sizes, is not possible due to flooding limitations. The catalysts need to be housed in special wire gauze envelopes as in the catalytic bales or KATAPAK-S configurations. A preliminary hardware design of a counter-current HDS reactor using catalytic bales was carried out in order to determine the technical feasibility. Using a realistic sulphur containing feedstock, the target of 50 ppm S content of desulphurized oil could be met in a reactor of reasonable dimensions. The study also underlines the need for accurate modelling of thermal effects during desulphurization. Our study also shows that interphase mass transfer is unlikely to be a limiting factor and there is a need to develop improved reactor configurations allowing for increased catalyst loading, at the expense of gas-liquid interfacial area. (Author)

  2. Accidents of loss of flow for the ETTR-2 reactor; deterministic analysis

    International Nuclear Information System (INIS)

    El-Messiry, A.M.

    2000-01-01

    The main objective for reactor safety is to keep the fuel in a thermally safe condition with adequate safety margins during all operational modes (normal-abnormal and accidental states). To achieve this purpose an accident analysis of different design base accident (DBA) as loss of flow accident (LOFA), is required assessing reactor safety. The present work concerns this transients applied to Egypt Test and Research Reactor ETRR-3 (new reactor). An accident analysis code FLOWTR is developed to investigate the thermal behaviour of the core during such flow transients. The active core is simulated by two channels: 1 - hot channel (HC), and 2 - average channel (AC) representing the remainder of the core. Each channel is divided into four axial sections. The external loop, core plenums, and core chimney are simulated by different dynamic loops. The code includes modules for pump cast down, flow regimes, decay heat, temperature distributions, and feedback coefficients. FLOWTR is verified against results from RETRAN code, THERMIC code and commissioning tests for null transient case. The comparison shows a good agreement. The study indicates that for LOFA transients, provided the scram system is available, the core is shutdown safely by low flow signal (496.6 kg/s) at 1.4 s were the HC temperature reaches the maximum value, 45.64 o C after shutdown. On the other hand, if the scram system is unavailable, and at t = 47.33 s, the core flow decreases to 67.41 kg/s, the HC temperature increases to 164.02 o C, and the HC clad surface heat flux exceeds its critical value of 400.00 W/cm 2 resulting of fuel burnout. (author)

  3. 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

  4. Reactor water level control device

    International Nuclear Information System (INIS)

    Utagawa, Kazuyuki.

    1993-01-01

    A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

  5. Decontamination and decommissioning project of the TRIGA Mark-2 and 3 research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jung, K J; Baik, S T; Chung, U S; Jung, K H; Park, S K; Lee, B J; Kim, J K; Yang, S H

    2000-01-01

    During the review on the decommissioning plan and environmental impact assessment report by the KINS, the number of the inquired items were two hundred and fifty one, and the answers were made and sent until September 10, 1999, as the screened review results were reported to Ministry of Science and Technology(MOST) in December 14, 1999, all the reviews on the licence were over. Radioactive liquid wastes of 400 tons generated during the operation of the research reactors including reactor vessels are stored in the facility of the research reactor 1 and 2. Those liquid wastes have the low-level-radioactivity which can be discharged to the surroundings, but was wholly treated to be vaporized naturally by means of the increased numbers of the natural vaporization disposal facilities with the annual capacity of 200 tons for the purpose of the minimized environmental contamination.

  6. Study on effects of development of reactor constant in fast reactor analysis

    International Nuclear Information System (INIS)

    Chiba, Gou

    2002-12-01

    Evaluation was carried out about an effect of development of the new generation reactor constant system that substitutes for the JFS library in fast reactor analysis. Analyzed cores were ZPPR in JUPITER critical experiment and several power reactor cores that were designed in the feasibility study. In the JUPITER analysis, large effects, over 10%, were observed in sodium void reactivity and sample Doppler reactivity. The former resulted from several factors, while the latter was due to an accurate of a resonance interaction effect between Doppler sample and core fuel. In the previous study, the effect had been evaluated in power reactor cores. The effect included an effect of corrosion of weighting spectrum because JFS-3-J3.2, which had been made with the incorrect weighting spectrum, was used in the evaluation. In the present study, JFS-3-J3.2R, which had been made with the correct weighting spectrum, was used. It was confirmed that the effect of development of reactor constant in power reactor was not as large as that in critical assembly. (author)

  7. Thermal and stress analyses of the reactor pressure vessel lower head of the Three Mile Island Unit 2

    International Nuclear Information System (INIS)

    Hashimoto, K.; Onizawa, K.; Kurihara, R.; Kawasaki, S.; Soda, K.

    1992-01-01

    Thermal and stress analyses were performed using the finite element analysis code ABAQUS to clarify the factors which caused tears in the stainless steel liner of the reactor pressure vessel lower head of the Three Mile Island Unit 2 (TMI-2) reactor pressure vessel during the accident on 28 March 1979. The present analyses covered the events which occurred after approximately 20 tons of molten core material were relocated to the lower head of the reactor pressure vessel. They showed that the tensile stress was highest in the case where the relocated core material consisting of homogeneous UO 2 debris was assumed to attack the lower head and the debris was then quenched. The peak tensile stress was in the vicinity of the welded zone of the penetration nozzle. This result agrees with the findings from the examination of the TMI-2 reactor pressure vessel that major tears in the stainless steel liner were observed around two penetration nozzles of the lower head. (author)

  8. The FR 2 reactor at Karlsruhe, F.R. Germany and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the FR 2 reactor and associated hot cell facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  9. Reactor physics measurements with 19-element ThOsub(2)-sup(235)UOsub(2) cluster fuel in heavy water moderator

    International Nuclear Information System (INIS)

    French, P.M.

    1985-02-01

    Low power lattice physics measurements have been performed with a single rod of 19-element thorium oxide fuel enriched with 1.45 wt. percent sub(235)UOsub(2) (93 percent enriched) in a simulated heavy water moderated and cooled power reactor core. The experiments were designed to provide data relevant to a power reactor irradiation and to obtain some basic information on the physics of uranium-thorium fuel material. Some theoretical flux calculations are summarized and show reasonable agreement with experiment

  10. The performance of ENDF/B-V.2 nuclear data for fast reactor calculations

    International Nuclear Information System (INIS)

    Atkinson, C.A.; Collins, P.J.

    1987-01-01

    Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations

  11. TMI-2 reactor vessel and balance of plant status

    International Nuclear Information System (INIS)

    Kuehn, G.A.

    1990-01-01

    In the fall of 1988 a corporate decision was made which concentrated effort on support of reactor vessel defueling and minimized activity in balance-of-plant areas. The auxiliary and fuel handling building are in a safe/stable state but final preparations for monitored storage won't be pursued until defueling and fuel shipping are complete. In addition to dispositioning fuel, the project is actively preparing for disposal of the Accident Generated Water (2.3 million gallons) by evaporation

  12. Well known and new neutron filters and employment of them for fundamental and applied investigations

    International Nuclear Information System (INIS)

    Gritzay, O.; Koloty, V.; Libman, V.; Kalchenko, O.; Klimova, N.

    2006-01-01

    This is a short review of the neutron filtered beam technique developed in Neutron Physics Department (NPD) at Kyiv Research Reactor (KRR) by now. The ways for search of new neutron filters and improving of known ones are described. A short information about characteristics of the existing filters is presented. This review may be useful for users who wish to use the existing filtered beams at KRR or to develop neutron filtered beam technique at their installations

  13. Abatement of fluorinated compounds using a 2.45 GHz microwave plasma torch with a reverse vortex plasma reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J.H.; Cho, C.H.; Shin, D.H. [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); Hong, Y.C., E-mail: ychong@nfri.re.kr [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); Shin, Y.W. [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); School of Advanced Green Energy and Environments, Handong Global University, Heunghae-eup, Buk-gu, Pohang-city, Gyeongbuk (Korea, Republic of)

    2015-08-30

    Highlights: • We developed a microwave plasma torch with reverse vortex reactor (RVR). • We calculated a volume fraction and temperature distribution of discharge gas and waste. • The performance of reverse vortex reactor increased from 29% to 43% than conventional vortex reactor. - Abstract: Abatement of fluorinated compounds (FCs) used in semiconductor and display industries has received an attention due to the increasingly stricter regulation on their emission. We have developed a 2.45 GHz microwave plasma torch with reverse vortex reactor (RVR). In order to design a reverse vortex plasma reactor, we calculated a volume fraction and temperature distribution of discharge gas and waste gas in RVR by ANSYS CFX of computational fluid dynamics (CFD) simulation code. Abatement experiments have been performed with respect to SF{sub 6}, NF{sub 3} by varying plasma power and N{sub 2} flow rates, and FCs concentration. Detailed experiments were conducted on the abatement of NF{sub 3} and SF{sub 6} in terms of destruction and removal efficiency (DRE) using Fourier transform infrared (FTIR). The DRE of 99.9% for NF{sub 3} was achieved without an additive gas at the N{sub 2} flow rate of 150 liter per minute (L/min) by applying a microwave power of 6 kW with RVR. Also, a DRE of SF{sub 6} was 99.99% at the N{sub 2} flow rate of 60 L/min using an applied microwave power of 6 kW. The performance of reverse vortex reactor increased about 43% of NF{sub 3} and 29% of SF{sub 6} abatements results definition by decomposition energy per liter more than conventional vortex reactor.

  14. Advances in reactor physics education: Visualization of reactor parameters

    International Nuclear Information System (INIS)

    Snoj, L.; Kromar, M.; Zerovnik, G.

    2012-01-01

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

  15. Slit-burst testing of cold-worked Zr-2.5 wt.% Nb pressure tubing for CANDU-PHW reactors

    International Nuclear Information System (INIS)

    Wilkins, B.J.S.; Barrie, J.N.; Zink, R.J.

    1978-12-01

    This report documents the available data on critical crack length of cold-worked Zr-2.5 wt.% Nb pressure tubing in CANDU reactors. In particular, it includes data for tubing removed from the Pickering 3 and 4 reactors. (author)

  16. Cronos 2: a neutronic simulation software for reactor core calculations; Cronos 2: un logiciel de simulation neutronique des coeurs de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Lautard, J J; Magnaud, C; Moreau, F; Baudron, A M [CEA Saclay, Dept. de Mecanique et de Technologie (DMT/SERMA), 91 - Gif-sur-Yvette (France)

    1999-07-01

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  17. TARMS, an on-line boiling water reactor operation management system. [3 D core simulator LOGOS 2

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, T.; Sakurai, S.; Uematsu, H.; Tsuiki, M.; Makino, K.

    1984-12-01

    The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site tool for boiling water reactor core operation management. It was designed to support a complete function set to meet the requirement to the current on-line process computers. The functions can be divided into two categories. One is monitoring of the present core power distribution as well as related limiting parameters. The other is aiding site engineers or reactor operators in making the future reactor operating plan. TARMS performs these functions with a three-dimensional BWR core physics simulator LOGOS 2, which is based on modified one-group, coarse-mesh nodal diffusion theory. A method was developed to obtain highly accurate nodal powers by coupling LOGOS 2 calculations with the readings of an in-core neutron flux monitor. A sort of automated machine-learning method also was developed to minimize the errors caused by insufficiency of the physics model adopted in LOGOS 2. In addition to these fundamental calculational methods, a number of core operation planning aid packages were developed and installed in TARMS, which were designed to make the operator's inputs simple and easy.

  18. Implementing new product development: A study of personal characteristics among managers

    DEFF Research Database (Denmark)

    Jespersen, Kristina Risom; Bysted, Rune

    2016-01-01

    Strategy implementation is an important aspect of innovation, as only 30% of planned strategies are matched by realised innovation outcomes. The present study investigates this in a setting of strategic implementation of New Product Development (NPD). In the implementation process, the managers...... strategic intentions create information overload in the NPD process, which decreases the new product novelty. In contrast, realisation of defensive strategic intentions makes experienced NPD managers focus on known NPD competences, thereby decreasing the new product novelty of NPD efforts. The study...

  19. Endoscopic management of pancreatic pseudocysts at atypical locations.

    Science.gov (United States)

    Bhasin, Deepak Kumar; Rana, Surinder Singh; Nanda, Mohit; Chandail, Vijant Singh; Masoodi, Ibrahim; Kang, Mandeep; Kalra, Navin; Sinha, Saroj Kant; Nagi, Birinder; Singh, Kartar

    2010-05-01

    There is paucity of data on endoscopic management of pseudocysts at atypical locations. We evaluated the efficacy of endoscopic transpapillary nasopancreatic drain (NPD) placement in the management of pseudocysts of pancreas at atypical locations. Eleven patients with pseudocysts at atypical locations were treated with attempted endoscopic transpapillary nasopancreatic drainage. On endoscopic retrograde pancreatography (ERP), a 5-F NPD was placed across/near the site of duct disruption. Three patients each had mediastinal, intrahepatic, and intra/perisplenic pseudocysts and one patient each had renal and pelvic pseudocyst. Nine patients had chronic pancreatitis whereas two patients had acute pancreatitis. The size of the pseudocysts ranged from 2 to 15 cm. On ERP, the site of ductal disruption was in the body of pancreas in five patients (45.4%), and tail of pancreas in six patients (54.6%). All the patients had partial disruption of pancreatic duct. The NPD was successfully placed across the disruption in 10 of the 11 patients (90.9%) and pseudocysts resolved in 4-8 weeks. One of the patients developed fever, 5 days after the procedure, which was successfully treated by intravenous antibiotics. In another patient, NPD became blocked 12 days after the procedure and was successfully opened by aspiration. The NPD slipped out in one of the patient with splenic pseudocyst and was replaced with a stent. There was no recurrence of symptoms or pseudocysts during follow-up of 3-70 months. Pancreatic pseudocysts at atypical locations with ductal communication and partial ductal disruption that is bridged by NPD can also be effectively treated with endoscopic transpapillary NPD placement.

  20. Reactor Core Internals Replacement of Ikata Units 1 and 2

    International Nuclear Information System (INIS)

    Ikeda, K.; Ishikawa, T.; Miyoshi, T.; Takagi, T.

    2012-01-01

    This paper presents an overview of the reactor core internals replacement project carried out at the Ikata Nuclear Power Station in Japan, which was the first of its kind among PWRs in the world. Failure of baffle former bolts was first reported in 1989 at Bugey 2 in France. Since then, similar incidents have been reported in Belgium and in the U.S., but not in Japan. However, the possibility of these bolts failing in Japanese plants cannot be denied in the future as operating hours increase. Ageing degradation mechanisms for the reactor core internals include irradiation-assisted stress corrosion cracking of baffle former bolts and mechanical wear of control rod guide cards. Two different approaches can be taken to address these ageing issues: to inspect and repair whenever a problem is found; and to replace the entire core internals with those of a new design having advanced features to prevent ageing degradation problems. The choice of our company was the latter. This paper explains the reasons for the choice and summarizes the replacement project activities at Ikata Units 1 and 2 as well as the improvements incorporated in the new design. (author)

  1. Investigation of the basic reactor physics characteristics 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; Khang, Ngo Phu [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    The Dalat nuclear research reactor was reconstructed from TRIGA MARK II reactor, built in 1963 with nominal power of 250 KW, and reached its planned nominal power of 500 kW for the first time in Feb. 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc. (author). 10 refs., 10 figs., 2 tabs.

  2. How Formal Control Influences Decision‐Making Clarity and Innovation Performance

    DEFF Research Database (Denmark)

    Schultz, Carsten; Salomo, Søren; de Brentani, Ulrike

    2013-01-01

    , which is a precise formal control mechanism relevant for managing specific problems at a single project level; (2) the immediate outcome of the application of formal controls, i.e. decision‐making clarity (DMC); and (3) degree of NPD innovativeness, a key contingency hypothesized to impact the efficacy...... of formal controls. For the empirical analysis, data are collected through a survey of 162 corporate NPD programs (Austria and Denmark, manufactured goods and services) where a total of 1274 respondents provide information relevant to their position. Hierarchical regression analysis is used to test...... the relationships. Results indicate that the performance effect of NPD formal control is fully mediated by DMC. Further, of the six hypothesized outcome relationships, four are fully supported. Both SGS and PM are effective systems for managing NPD when degree of innovativeness is not taken into account. PM...

  3. 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.)

  4. Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

    International Nuclear Information System (INIS)

    D. Kokkinos

    2005-01-01

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

  5. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Kalcheva, S [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Sikik, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-09-01

    BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).

  6. 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.

  7. Narcissistic Personality Disorder and the Structure of Common Mental Disorders.

    Science.gov (United States)

    Eaton, Nicholas R; Rodriguez-Seijas, Craig; Krueger, Robert F; Campbell, W Keith; Grant, Bridget F; Hasin, Deborah S

    2017-08-01

    Narcissistic personality disorder (NPD) shows high rates of comorbidity with mood, anxiety, substance use, and other personality disorders. Previous bivariate comorbidity investigations have left NPD multivariate comorbidity patterns poorly understood. Structural psychopathology research suggests that two transdiagnostic factors, internalizing (with distress and fear subfactors) and externalizing, account for comorbidity among common mental disorders. NPD has rarely been evaluated within this framework, with studies producing equivocal results. We investigated how NPD related to other mental disorders in the internalizing-externalizing model using diagnoses from a nationally representative sample (N = 34,653). NPD was best conceptualized as a distress disorder. NPD variance accounted for by transdiagnostic factors was modest, suggesting its variance is largely unique in the context of other common mental disorders. Results clarify NPD multivariate comorbidity, suggest avenues for classification and clinical endeavors, and highlight the need to understand vulnerable and grandiose narcissism subtypes' comorbidity patterns and structural relations.

  8. Applications: fission, nuclear reactors. Fission: the various ways for reactors and cycles

    International Nuclear Information System (INIS)

    Bacher, P.

    1997-01-01

    A historical review is presented concerning the various nuclear reactor systems developed in France by the CEA: the UNGG (graphite-gas) system with higher CO 2 pressures, bigger fuel assemblies and powers higher than 500 MW e, allowed by studies on reactor physics, cladding material developments and reactor optimization; the fast neutron reactor system, following the graphite-gas development, led to the Superphenix reactor and important progress in simulation based on experiment and return of experience; and the PWR system, based on the american license, which has been successfully accommodated to the french industry and generates up to 75% of the electric power in France

  9. A method of reactor power decrease by 2DOF control system during BWR power oscillation

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Suzuki, Katsuo

    1998-09-01

    Occurrence of power oscillation events caused by void feedback effects in BWRs operated at low-flow and high-power condition has been reported. After thoroughly examining these events, BWRs have been equipped with the SRI (Selected Rod Insertion) system to avoid the power oscillation by decreasing the power under such reactor condition. This report presents a power control method for decreasing the reactor power stably by a two degree of freedom (2DOF) control. Performing a numerical simulation by utilizing a simple reactor dynamics model, it is found that the control system designed attains a satisfactory control performance of power decrease from a viewpoint of setting time and oscillation. (author)

  10. 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.)

  11. 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.

  12. Catalytic combustion of the retentate gas from a CO2/H2 separation membrane reactor for further CO2 enrichment and energy recovery

    International Nuclear Information System (INIS)

    Hwang, Kyung-Ran; Park, Jin-Woo; Lee, Sung-Wook; Hong, Sungkook; Lee, Chun-Boo; Oh, Duck-Kyu; Jin, Min-Ho; Lee, Dong-Wook; Park, Jong-Soo

    2015-01-01

    The CCR (catalytic combustion reaction) of the retentate gas, consisting of 90% CO 2 and 10% H 2 obtained from a CO 2 /H 2 separation membrane reactor, was investigated using a porous Ni metal catalyst in order to recover energy and further enrich CO 2 . A disc-shaped porous Ni metal catalyst, namely Al[0.1]/Ni, was prepared by a simple method and a compact MCR (micro-channel reactor) equipped with a catalyst plate was designed for the CCR. CO 2 and H 2 concentrations of 98.68% and 0.46%, respectively, were achieved at an operating temperature of 400 °C, GHSV (gas-hourly space velocity) of 50,000 h −1 and a H 2 /O 2 ratio (R/O) of 2 in the unit module. In the case of the MCR, a sheet of the Ni metal catalyst was easily installed along with the other metal plates and the concentration of CO 2 in the retentate gas increased up to 96.7%. The differences in temperatures measured before and after the CCR were 31 °C at the product outlet and 19 °C at the N 2 outlet in the MCR. The disc-shaped porous metal catalyst and MCR configuration used in this study exhibit potential advantages, such as high thermal transfer resulting in improved energy recovery rate, simple catalyst preparation, and easy installation of the catalyst in the MCR. - Highlights: • The catalytic combustion of a retentate gas obtained from the H 2 /CO 2 separation membrane. • A disc-shaped porous nickel metal catalyst and a micro-channel reactor for catalytic hydrogen combustion. • CO 2 enrichment up to 98.68% at 400 °C, 50,000 h −1 and H 2 /O 2 ratio of 2.

  13. Reactor feedwater system

    International Nuclear Information System (INIS)

    Kagaya, Hiroyuki; Tominaga, Kenji.

    1993-01-01

    In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

  14. Information paths within the new product development process

    DEFF Research Database (Denmark)

    Jespersen, Kristina Risom

    2007-01-01

    collection platform to obtain measurements from within the NPD process. 42 large, international companies participated in the data collecting simulation. Results revealed five different information paths that were not connecting all stages of the NPD process. Moreover, results show that the front......-end is not driving the information acquisition through the stages of the NPD process, and that environmental turbulence disconnects stages from the information paths in the NPD process. This implies that information is at the same time a key to success and a key to entrapment in the NPD process....

  15. Examining sex differences in DSM-IV-TR narcissistic personality disorder symptom expression using Item Response Theory (IRT).

    Science.gov (United States)

    Hoertel, Nicolas; Peyre, Hugo; Lavaud, Pierre; Blanco, Carlos; Guerin-Langlois, Christophe; René, Margaux; Schuster, Jean-Pierre; Lemogne, Cédric; Delorme, Richard; Limosin, Frédéric

    2017-12-14

    The limited published literature on the subject suggests that there may be differences in how females and males experience narcissistic personality disorder (NPD) symptoms. The aim of this study was to use methods based on item response theory to examine whether, when equating for levels of NPD symptom severity, there are sex differences in the likelihood of reporting DSM-IV-TR NPD symptoms. We conducted these analyses using a large, nationally representative sample from the USA (n=34,653), the second wave of the National Epidemiologic Survey on Alcohol and Related Conditions (NESARC). There were statistically and clinically significant sex differences for 2 out of the 9 DSM-IV-TR NPD symptoms. We found that males were more likely to endorse the item 'lack of empathy' at lower levels of narcissistic personality disorder severity than females. The item 'being envious' was a better indicator of NPD severity in males than in females. There were no clinically significant sex differences on the remaining NPD symptoms. Overall, our findings indicate substantial sex differences in narcissistic personality disorder symptom expression. Although our results may reflect sex-bias in diagnostic criteria, they are consistent with recent views suggesting that narcissistic personality disorder may be underpinned by shared and sex-specific mechanisms. Copyright © 2017 Elsevier B.V. All rights reserved.

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

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  17. 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)

  18. Model study of an automatic controller of the IBR-2 pulsed reactor

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Popov, A.K.

    2007-01-01

    For calculation of power transients in the IBR-2 reactor a special mathematical model of dynamics taking into account the discontinuous jump of reactivity by an automatic controller with the step motor is created. In the model the nonlinear dependence of the energy of power pulse on the reactivity and the influence of warming up of the reactor on the reactivity by means of introduction of a nonlinear feedback 'power-pulse energy - reactivity' are taken into account. With the help of the model the transients of relative deviation of power-pulse energy are calculated at various (random, mixed and regular) reactivity disturbances at the reactor mean power 1.475 MW. It is shown that to improve the quality of processes the choice of such regular values of parameters of the automatic controller is expedient, at which the least effect of smoothing of a signal acting on an automatic controller and the least speed of an automatic controller are provided, and the reduction of efficiency of one step of the automatic controller and introduction of a five-percent dead space are also expedient

  19. Backfitting of the FRG reactors

    Energy Technology Data Exchange (ETDEWEB)

    Krull, W [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht (Germany)

    1990-05-01

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U{sub 3}Si{sub 2} fuel. Both cooling towers were repaired. Replacement of instrumentation is planned.

  20. Backfitting of the FRG reactors

    International Nuclear Information System (INIS)

    Krull, W.

    1990-01-01

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U 3 Si 2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

  1. 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)

  2. Design options for a bunsen reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Robert Charles

    2013-10-01

    This work is being performed for Matt Channon Consulting as part of the Sandia National Laboratories New Mexico Small Business Assistance Program (NMSBA). Matt Channon Consulting has requested Sandia's assistance in the design of a chemical Bunsen reactor for the reaction of SO2, I2 and H2O to produce H2SO4 and HI with a SO2 feed rate to the reactor of 50 kg/hour. Based on this value, an assumed reactor efficiency of 33%, and kinetic data from the literature, a plug flow reactor approximately 1%E2%80%9D diameter and and 12 inches long would be needed to meet the specification of the project. Because the Bunsen reaction is exothermic, heat in the amount of approximately 128,000 kJ/hr would need to be removed using a cooling jacket placed around the tubular reactor. The available literature information on Bunsen reactor design and operation, certain support equipment needed for process operation and a design that meet the specification of Matt Channon Consulting are presented.

  3. Relative neutronic performance of proposed high-density dispersion fuels in water-moderated and D2O-reflected research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Matos, J.E.; Snelgrove, J.L.

    1996-01-01

    This paper provides an overview of the neutronic performance of an idealized research reactor using several high density LEU fuels that are being developed by the RERTR program. High-density LEU dispersion fuels are needed for new and existing high-performance research reactors and to extend the lifetime of fuel elements in other research reactors. This paper discusses the anticipated neutronic behavior of proposed advanced fuels containing dispersions of U 3 Si 2 , UN, U 2 Mo and several uranium alloys with Mo, or Zr and Nb. These advanced fuels are ranked based on the results of equilibrium depletion calculations for a simplified reactor model having a small H 2 O-cooled core and a D 2 O reflector. Plans have been developed to fabricate and irradiate several uranium alloy dispersion fuels in order to test their stability and compatibility with the matrix material and to establish practical loading limits

  4. Relative neutronic performance of proposed high-density dispersion fuels in water-moderated and D2O-reflected research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Matos, J.E.; Snelgrove, J.L.

    1996-01-01

    This paper provides an overview of the neutronic performance of an idealized research reactor using several high density Leu fuels that are being developed by the Rarita program. High-density Leu dispersion fuels are needed for new and existing high-performance research reactors and to extend the lifetime of fuel elements in other research reactors. This paper discusses the anticipated neutronic behavior of proposed advanced fuels containing dispersions of U 3 Si 2 , UN, U 2 Mo and several uranium alloys with Mo, or Zr and Nb. These advanced fuels are ranked based on the results of equilibrium depletion calculations for a simplified reactor model having a small H 2 O-cooled core and a D 2 O reflector. Plans have been developed to fabricate and irradiate several uranium alloy dispersion fuels in order to test their stability and compatibility with the matrix material and to establish practical loading limits. (author)

  5. Performance analysis of photocatalytic CO2 reduction in optical fiber monolith reactor with multiple inverse lights

    International Nuclear Information System (INIS)

    Yuan, Kai; Yang, Lijun; Du, Xiaoze; Yang, Yongping

    2014-01-01

    Highlights: • A new optical fiber monolith reactor model for CO 2 reduction was developed. • Methanol concentration versus fiber location and operation parameters was obtained. • Reaction efficiency increases by 31.1% due to the four fibers and inverse layout. • With increasing space of fiber and channel center, methanol concentration increases. • Methanol concentration increases as the vapor ratio and light intensity increase. - Abstract: Photocatalytic CO 2 reduction seems potential to mitigate greenhouse gas emissions and produce renewable energy. A new model of photocatalytic CO 2 reduction in optical fiber monolith reactor with multiple inverse lights was developed in this study to improve the conversion of CO 2 to CH 3 OH. The new light distribution equation was derived, by which the light distribution was modeled and analyzed. The variations of CH 3 OH concentration with the fiber location and operation parameters were obtained by means of numerical simulation. The results show that the outlet CH 3 OH concentration is 31.1% higher than the previous model, which is attributed to the four fibers and inverse layout. With the increase of the distance between the fiber and the monolith center, the average CH 3 OH concentration increases. The average CH 3 OH concentration also rises as the light input and water vapor percentage increase, but declines with increasing the inlet velocity. The maximum conversion rate and quantum efficiency in the model are 0.235 μmol g −1 h −1 and 0.0177% respectively, both higher than previous internally illuminated monolith reactor (0.16 μmol g −1 h −1 and 0.012%). The optical fiber monolith reactor layout with multiple inverse lights is recommended in the design of photocatalytic reactor of CO 2 reduction

  6. MASTER-2.0: Multi-purpose analyzer for static and transient effects of reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Byung Oh; Song, Jae Seung; Joo, Han Gyu [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    MASTER-2.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the two group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM(Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with AFEN/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. Master-2.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P model can be used selectively. In addition, MASTER-2.0 is designed to cover various PWRs including SMART as well as WH-and CE-type reactors, providing all data required in their design procedures. (author). 39 refs., 12 figs., 4 tabs.

  7. Fast reactors worldwide

    International Nuclear Information System (INIS)

    Hall, R.S.; Vignon, D.

    1985-01-01

    The paper concerns the evolution of fast reactors over the past 30 years, and their present status. Fast reactor development in different countries is described, and the present position, with emphasis on cost reduction and collaboration, is examined. The French development of the fast breeder type reactor is reviewed, and includes: the acquisition of technical skills, the search for competitive costs and the spx2 project, and more advanced designs. Future prospects are also discussed. (U.K.)

  8. Application of stable adaptive schemes to nuclear reactor systems, (2)

    International Nuclear Information System (INIS)

    Kukuda, Toshio

    1979-01-01

    The parameter identification and adaptive control schemes applied in a previous study to a nonlinear point reactor are extended to the case of a loosely-coupled-core reactor with internal feedbacks, constituting a nonlinear overall system. Both schemes are shown to be stable, with the system newly represented on the pattern of the Model Reference Adaptive System (MRAS) with use made of the Lyapunov's method. For either parameter identification or adaptive control of a loosely-coupled-core reactor, there exists no canonical form of multiple input-multiple output system which can be directly applied for deriving the MRAS with the matrix version of the Kalman-Yakubovich lemma as it was in the case of the point reactor. This difficulty is circumvented by the practical assumption that the neutron density can be directly measured on each core as reactivity change is applied as input into the coupled core as a whole. For parameter identification, the model parameters are adaptively adjusted to those of each core, while for the adaptive control, plant parameters of each core can be adaptively compensated, again through control inputs, to asymptotically reduce the output error between the model and the plant. The point reactor is shown to correspond to a special case. (author)

  9. Upgradation of Apsara reactor

    International Nuclear Information System (INIS)

    Mammen, S.; Mukherjee, P.; Bhatnagar, A.; Sasidharan, K.; Raina, V.K.

    2009-01-01

    Apsara is a 1 MW swimming pool type research reactor using high enriched uranium as fuel with light water as coolant and moderator. The reactor is in operation for more than five decades and has been extensively used for basic research, radioisotope production, neutron radiography, detector testing, shielding experiments etc. In view of its long service period, it is planned to carry out refurbishment of the reactor to extend its useful life. During refurbishment, it is also planned to upgrade the reactor to a 2 MW reactor to improve its utilization and to upgrade the structure, system and components in line with the current safety standards. This paper gives a brief account of the design features and safety aspects of the upgraded Apsara reactor. (author)

  10. The use of ICT tools to support collaborative product development activities: evidences from Brazilian industry

    Directory of Open Access Journals (Sweden)

    Daisy Valle Enrique

    2018-06-01

    Full Text Available Abstract Paper aims This paper aims to understand the relationship between Information & Communication Technology (ICT, collaborative New Product Development (NPD and customer satisfaction (NPD performance. Originality We target the relationship between ICT, collaborative NPD and NPD performance. ICT is assessed as a set of specific tools adopted by the companies. Research method We test the mediating role of collaborative practices in the effect of ICT tools on customer satisfaction (as NPD performance by means of a survey of 105 Brazilian firms. Main findings Collaboration with customers and suppliers has an important role for customer satisfaction and the use of ICT has significant effect on NPD performance through the mediating role of collaborative practices. Implications for theory and practice Implementing only ICT tools is not enough to achieve higher level of success in NPD. Managers should first strength the relationship between stakeholders and then adopt ICT tools to support the cooperation.

  11. Mixed core management: Use of 93% and 72% enriched uranium in the BR2 reactor

    International Nuclear Information System (INIS)

    Ponsard, B.

    2000-01-01

    The BR2 reactor, put into operation in 1963 and refurbished from July 1995 till April 1997, is a 100 MW high-flux Materials Testing Reactor, using 93% 235 U enriched uranium as standard fuel, light water as coolant and beryllium as moderator. The present operating regime consists of five irradiation cycles per year at an operating power between 50 and 70 MW; each cycle is characterized by 21 days operation. In the framework of a 'qualification programme', six 72% 235 U fuel elements fabricated with uranium recovered from the reprocessing of BR2 spent fuel at UKAEA-Dounreay have been successfully irradiated in the period 1994-1995 reaching a maximum mean burnup of 48% without the release of fission products. Since 1998, this type of fuel element is irradiated routinely together with standard 93% 235 U fuel elements in order to optimize the utilization of the available HEU inventory. The purpose of this paper is to present the strategy developed in order to optimize the mixed core management of the BR2 reactor. (author)

  12. Effect of fuel assembly when changing from AFA 2G to AFA 3G on seismic loads of reactor internal

    International Nuclear Information System (INIS)

    Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

    2013-01-01

    Nonlinear seismic model for reactor with fuel assemblies of AFA 2G and AFA 3G is established. Using ANSYS software, seismic nonlinear time -history analysis is completed and the effects on seismic loads of reactor system are obtained. The result shows that when the fuel assembly changing from AFA 2G to AFA 3G, it is necessary to reevaluate the fuel assembly itself, but not the reactor internal. (authors)

  13. Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas V, G.; Banfield, J. [GE-Hitachi Nuclear Energy Americas LLC, Global Nuclear Fuel, Americas LLC, 3901 Castle Hayne Road, Wilmingtonm, North Carolina (United States); Avila N, A., E-mail: Gabriel.Cuevas-Vivas@ge.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

    2016-09-15

    This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

  14. Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

    International Nuclear Information System (INIS)

    Cuevas V, G.; Banfield, J.; Avila N, A.

    2016-09-01

    This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

  15. Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

    International Nuclear Information System (INIS)

    González Mantecón, Javier

    2015-01-01

    The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

  16. CO2 Energy Reactor - Integrated Mineral Carbonation: Perspectives on Lab-Scale Investigation and Products Valorization

    OpenAIRE

    Rafael M Santos; Pol CM Knops; Keesjan L Rijnsburger; Yi Wai eChiang

    2016-01-01

    To overcome the challenges of mineral CO2 sequestration, Innovation Concepts B.V. is developing a unique proprietary gravity pressure vessel (GPV) reactor technology and has focussed on generating reaction products of high economic value. The GPV provides intense process conditions through hydrostatic pressurization and heat exchange integration that harvests exothermic reaction energy, thereby reducing energy demand of conventional reactor designs, in addition to offering other benefits. In ...

  17. What occurred in the reactors

    International Nuclear Information System (INIS)

    Kudo, Kazuhiko

    2013-01-01

    Described is what occurred in the reactors of Fukushima Daiichi Nuclear Power Plant at the Tohoku earthquake and tsunami (Mar. 11, 2011) from the aspect of engineering science. The tsunami attacked the Plant 1 hr after the quake. The Plant had reactors in buildings no.1-4 at 10 m height from the normal sea level which was flooded by 1.5-5.5 m high wave. All reactors in no.1-6 in the Plant were the boiling water type, and their core nuclear reactions were stopped within 3 sec due to the first quake by control rods inserted automatically. Reactors in no.1-5 lost their external AC power sources by the breakdown and subsequent submergence (no.1-4) of various equipments and in no.1, 2 and 4, the secondary DC power was then lost by the battery death. Although the isolation condenser started to cool the reactor in no.1 after DC cut, its valve was then kept closed to heat up the reactor, leading to the reaction of heated Zr in the fuel tube and water to yield H 2 which was accumulated in the building: the cause of hydrogen explosion on 12th. The reactor in no.2 had the reactor core isolation cooling system (RCIC) which operated normally for few hrs, then probably stopped to heat up the reactor, resulting in meltdown of the core but no explosion occurred because of the opened door of the blowout panel on the wall by the blast of no.1 explosion. The reactor in no.3 had RCIC and high pressure coolant injection system, but their works stopped to result in the core damage and H 2 accumulation leading to the explosion on 14th. The reactor in no.4 had not been operated because of its periodical annual examination, but was explored on 15th, of which cause was thought to be due to backward flow of H 2 from no.3. Finally, the author discusses about this accident from the industrial aspect of the design of safety level (defense in depth) on international views, and problems and tasks given. (T.T.)

  18. Disturbed emotion recognition in patients with narcissistic personality disorder.

    Science.gov (United States)

    Marissen, Marlies A E; Deen, Mathijs L; Franken, Ingmar H A

    2012-07-30

    Although theoretically the lack of empathy is a supposed key symptom of narcissistic personality disorder (NPD), empirical studies examining empathy in NPD are scarce. In the present study it was examined whether patients with NPD differ from healthy controls and a psychiatric control group in their empathic abilities. In order to examine this question, 20 patients with NPD, 20 patients with a personality disorder in the Cluster C spectrum and 20 healthy control participants were presented with a questionnaire and a facial recognition task designed to measure empathic abilities. It was found that patients with NPD did not differ from the two control groups on a self-report questionnaire indicating that patients regard themselves as sensitive to the feelings of others. On the contrary, it was found NPD patients generally performed worse on a facial emotion recognition task compared to both control groups. In addition to this general deficit in emotion recognition, patients with NPD showed a specific deficit for emotions representing fear and disgust. These results provide the first empirical evidence for impaired emotion recognition in patients with NPD. Crown Copyright © 2012. Published by Elsevier Ireland Ltd. All rights reserved.

  19. 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.)

  20. Catalytic wet oxidation of phenol in a trickle bed reactor over a Pt/TiO2 catalyst.

    Science.gov (United States)

    Maugans, Clayton B; Akgerman, Aydin

    2003-01-01

    Catalytic wet oxidation of phenol was studied in a batch and a trickle bed reactor using 4.45% Pt/TiO2 catalyst in the temperature range 150-205 degrees C. Kinetic data were obtained from batch reactor studies and used to model the reaction kinetics for phenol disappearance and for total organic carbon disappearance. Trickle bed experiments were then performed to generate data from a heterogeneous flow reactor. Catalyst deactivation was observed in the trickle bed reactor, although the exact cause was not determined. Deactivation was observed to linearly increase with the cumulative amount of phenol that had passed over the catalyst bed. Trickle bed reactor modeling was performed using a three-phase heterogeneous model. Model parameters were determined from literature correlations, batch derived kinetic data, and trickle bed derived catalyst deactivation data. The model equations were solved using orthogonal collocations on finite elements. Trickle bed performance was successfully predicted using the batch derived kinetic model and the three-phase reactor model. Thus, using the kinetics determined from limited data in the batch mode, it is possible to predict continuous flow multiphase reactor performance.

  1. [Comorbidity in patients with narcissistic personality disorder in comparison to patients with borderline personality disorder].

    Science.gov (United States)

    Ritter, Kathrin; Roepke, Stefan; Merkl, Angela; Heuser, Isabella; Fydrich, Thomas; Lammers, Claas-Hinrich

    2010-01-01

    Patients with a narcissistic personality disorder (NPD) do not often consult a psychotherapist or psychiatrist because of their NPD, but rather, because of co-occurring psychiatric disorders, or higher general symptom stress. Until now there is no actual data about rates of co-occurrence disorders and general symptom stress. Which axis I and axis II disorders occur typically in NPD in comparison to patients with a borderline personality disorder (BPD)? How are general symptom stress and depressive symptoms related? Prevalence of co-occurring disorders (Structured Clinical Interview for DSM-IV for Axis I and Axis II) and general symptom stress (SCL-90-R) and depression (BDI) were investigated in 62 patients with a NPD, 62 patients with a BPD and 59 patients with a double diagnosis NPD/BPD. Affective disorders (64.5%) and substance use disorders (35.5%) were the most comorbid psychiatric disorders in patients with NPD. Substance use disorders (pdisorder (PTSD) (pdisorders (ppersonality disorder (pdisorders and antisocial personality disorder. Patients with NPD showed lowest rates of co-occurring disorders and lowest scores in general symptom stress and depression than the other two groups. In general, patients with NPD showed similar co-occurring disorders as patients with BPD, or with the co-diagnosis NPD and BPD, but they showed lower scores for general symptom stress and depression. (c) Georg Thieme Verlag KG Stuttgart New York.

  2. Investigation of sensors and instrument components in boiling water reactors. Results from Oskarshamn 2, Barsebaeck 2 in Sweden and Kernkraftwerk Muehleberg in Switzerland

    International Nuclear Information System (INIS)

    Bergdahl, B.G.

    1998-05-01

    The reactor monitoring instruments are important for the operation and safety of the plants. Static properties of the instruments are controlled annually, but the dynamic properties are rarely, if ever, examined. This study is the result of a project initiated by the Swedish Nuclear Power Inspectorate. The examinations are based on signal analysis and simultaneous measurement of multiple signals. Results from Oskarshamn 2 (O2), Barsebaeck 2 (B2) and Kernkraftwerk Muehleberg (KKM) are discussed in this report. The presentation is focused on reactor pressure and reactor level signals. the analysis of O2 revealed that the dynamics for 3 out of 14 sensors was 'filtered', meaning that a rapid level displacement is registered with delay. Inspection showed that a 1 sec filter was installed instead of 1.2 sec. The study also showed that old pressure-sensors in use both at O2 and B2 could not cope with high frequencies, and that some level-sensors were disturbed by mechanical oscillations at Bw. At KKM, a 2 Hz resonance was observed with 12 pressure and level sensors. The oscillation was created by an old pressure sensor and influenced the other sensors through the common impulse network

  3. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    International Nuclear Information System (INIS)

    Guigon, B.; Vacelet, H.; Dornbusch, D.

    2000-01-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs from activation analysis to power reactor fuel qualification. In this paper the main characteristics of the Jules Horowitz Reactor are presented. Safety criteria are explained. Finally, merits and disadvantages of UMo compared to the standard U 3 Si 2 fuel are discussed. (author)

  4. TMI-2 [Three Mile Island Unit 2] reactor building dose reduction task force

    International Nuclear Information System (INIS)

    Daniels, R.S.

    1988-01-01

    In late October 1982, the director of Three Mile Island Unit 2 (TMI-2) created the dose reduction task force with the objective of identifying the principal radiological sources in the reactor building and recommending actions to minimize the dose to workers on labor-intensive projects. Members of the task force were drawn form various groups at TMI. Findings and recommendations were presented to the US Nuclear Regulatory Commission in a briefing on November 18, 1982. The task force developed a three-step approach toward dose reduction. Step 1 identified the radiological sources. Step 2 modeled the source and estimated its contribution to the general area dose rates. Step 3 recommended actions to achieve dose reductions consistent with general exposure rate goals

  5. A novel auto-thermal reforming membrane reactor for high purity H2

    International Nuclear Information System (INIS)

    Tony Boyd; Grace, J.R.; Lim, C.J.; Adris, A.M.

    2006-01-01

    A novel hydrogen reactor based on steam reforming of natural gas has been developed and tested. The reactor produces high purity hydrogen using in-situ perm-selective membranes installed in a fluidized catalyst bed, thus shifting the thermodynamic equilibrium of the SMR reaction and eliminating the need for downstream hydrogen purification. The reactor is particularly suited to auto-thermal reforming, where air is added to the reformer to provide the endothermic reaction heat, thus eliminating the need to indirectly heat the reactor. The gas flow pattern within the fluidized bed induces an internal circulation of catalyst particles between the central SMR reaction (permeation) zone and an outer annulus. The circulating hot catalyst particles from the oxidation zone carry the required endothermic heat of reaction for the reforming, while ensuring that the palladium membranes are not exposed to excessive temperatures or to oxygen. Another beneficial characteristic of the reactor is that very little of the nitrogen present in the oxidation air reaches the reaction zone, thus maintaining the hydrogen driving force for the perm-selective membranes. Pilot plant results carried out in a semi-industrial scale reactor will be presented. The reactor was operated up to 650 C and 14 bar. Pure hydrogen (99.999+%) was initially obtained from the reactor and an equilibrium shift was demonstrated. (authors)

  6. Operating US power reactors

    International Nuclear Information System (INIS)

    Silver, E.G.

    1988-01-01

    This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of September 30, 1987, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (July, August, and September 1987) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. In addition to the tabular data, this article discusses other significant occurrences and developments that affected licensed US power reactors during this reporting period. Status changes at Braidwood Unit 1, Nine Mile Point 2, and Beaver Valley 2 are discussed. Other occurrences discussed are: retraining of control-room operators at Peach Bottom; a request for 25% power for Shoreham, problems at Fermi 2 which delayed the request to go to 75% power; the results of a safety study of the N Reactor at Hanford; a proposed merger of Pacific Gas and Electric with Sacramento Municipal Utility District which would result in the decommissioning of Rancho Seco; the ordered shutdown of Oyster Creek; a minor radioactivity release caused by a steam generator tube rupture at North Anna 1; and 13 fines levied by the NRC on reactor licensees

  7. Necessity of research reactors

    International Nuclear Information System (INIS)

    Ito, Tetsuo

    2016-01-01

    Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

  8. The validity of the Personality Diagnostic Questionnaire-4 Narcissistic Personality Disorder scale for assessing pathological grandiosity.

    Science.gov (United States)

    Hopwood, Christopher J; Donnellan, M Brent; Ackerman, Robert A; Thomas, Katherine M; Morey, Leslie C; Skodol, Andrew E

    2013-01-01

    Although controversy surrounds the definition and measurement of narcissism, the claim that pathological grandiosity is central to the construct generates little disagreement. Yet representations of pathological grandiosity vary across measures of narcissism, leading to conceptual confusion in the literature. The validity of a DSM-based measure of pathological narcissism, the Personality Diagnostic Questionnaire-4 Narcissistic Personality Disorder scale (PDQ-4 NPD), was evaluated in 1 clinical and 3 nonclinical samples (total N=2,391) for its ability to measure pathological grandiosity. Findings were generally supportive: average scores were higher in the clinical than nonclinical samples and the PDQ-4 NPD scale correlated most strongly with (a) other measures of NPD; (b) other DSM Cluster B personality disorders; (c) traits involving antagonism, hostility, and assertiveness; and (d) interpersonal distress and disaffiliative dominance. However, the low internal consistency of the PDQ-4 NPD scale and unexpected associations with Cluster A and obsessive-compulsive features point to potential psychometric weaknesses with this instrument. These findings are useful for evaluating the PDQ-4 NPD scale and for informing ongoing debates regarding how to define and assess pathological narcissism.

  9. 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

  10. Environmental assessment for decontamination of the Three Mile Island Unit 2 reactor building atmosphere. Addendum 2. Draft NRC staff report for public comment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-04-01

    The reactor building purge system is an existing system originally installed for purging the reactor building atmosphere during normal operation or maintenance conditions. Use of the reactor building purge system in conjunction with the hydrogen control subsystem evaluated in Section 6.1 represents a variation in the purging alternative for decontaminating the Unit 2 reactor building atmosphere. This variation in the purging alternative would function only under meteorological conditions favorable to atmospheric dispersion. The reactor building purge system is capable of purging the building at flow rates of 5,000-50,000 cfm. Actual purge rates authorized during any time interval would be dependent on meteorological conditions and reactor building concentrations. Like the hydrogen control subsystem, this system would remove reactor building atmosphere through a filter system and discharge it through the 160-ft plant vent stack to the environment. The advantage of using the reactor building purge system in conjunction with the hydrogen control system is that it could decontaminate the reactor building atmosphere in a total elapsed purge time as short as approximately 5 days, as compared with the 60 days that would be required if the hydrogen purge subsystem were used alone. Use of this variation in the purge alternative would result in the release of radioactive materials to the environment. However, calculations based on actual meteorological and release-rate data would be used to monitor radioactive releases so that they do not exceed the requirements of 10 CFR Part 20, the design objectives of 10 CFR Part 50, Appendix 1 and the applicable requirements of 40 CFR 190.10.

  11. Re criticality assessment following reactor core damage in Fukushima unit 2

    International Nuclear Information System (INIS)

    Jeong, Hae Sun; Song, Jin Ho; Park, Chang Je; Ha, Kwang Soon; Song, Yong Mann; Ryu, Eun Hyun

    2012-01-01

    Following the severe core damage accident at the Fukushima nuclear power plants (NPPs), many researchers have studied a possible Re criticality caused by core melting or corium. However, no one can accurately examine the internal conditions of the reactor vessel, and thus there have been different opinions from some organizations depending on their assumption and analysis methods. If there is a potential Re criticality in the reactor vessel, some counter plans for the accident management should be established to prevent and mitigate re criticality, and to return the plant to a safe and stable state. In this study, the criticality level following a severe core damage accident was first analyzed using the MCNPX 2.6.0 code. Based on this result, practical strategies in terms of accident management were obtained by charging soluble boron (H 3B O 3) into re flooded water

  12. RB reactor as the U-D2O benchmark criticality system

    International Nuclear Information System (INIS)

    Pesic, M.

    1998-01-01

    From a rich and valuable database fro 580 different reactor cores formed up to now in the RB nuclear reactor, a selected and well recorded set is carefully chosen and preliminarily proposed as a new uranium-heavy water benchmark criticality system for validation od reactor design computer codes and data libraries. The first results of validation of the MCNP code and adjoining neutron cross section libraries are resented in this paper. (author)

  13. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 2. User's manual

    International Nuclear Information System (INIS)

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear energy reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 2: User's Manual) describes the input requirements of VIPRE and its auxiliary programs, SPECSET, ASP and DECCON, and lists the input instructions for each code

  14. Cadmium-emitter self-powered thermal neutron detector performance characterization & reactor power tracking capability experiments performed in ZED-2

    Energy Technology Data Exchange (ETDEWEB)

    LaFontaine, M.W., E-mail: physics@execulink.com [LaFontaine Consulting, Kitchener, Ontario (Canada); Zeller, M.B. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Nielsen, K. [Royal Military College of Canada, SLOWPOKE-2 Reactor, Kingston, Ontario (Canada)

    2014-07-01

    Cadmium-emitter self-powered thermal neutron flux detectors (SPDs), are typically used for flux monitoring and control applications in low temperature, test reactors such as the SLOWPOKE-2. A collaborative program between Atomic Energy of Canada, academia (Royal Military College of Canada (RMCC)) and industry (LaFontaine Consulting) was initiated to characterize the incore performance of a typical Cd-emitter SPD; and to obtain a definitive measure of the capability of the detector to track changes in reactor power in real time. Prior to starting the experiment proper, Chalk River Laboratories' ZED-2 was operated at low power (5 watts nominal) to verify the predicted moderator critical height. Test measurements were then performed with the vertical center of the SPD emitter positioned at the vertical mid-plane of the ZED-2 reactor core. Measurements were taken with the SPD located at lattice position L0 (near center), and repeated at lattice position P0 (in D{sub 2}O reflector). An ionization chamber (part of the ZED-2 control instrumentation) monitored reactor power at a position located on the south side of the outside wall of the reactor's calandria. These experiments facilitated measurement of the absolute thermal neutron sensitivity of the subject Cd-emitter SPD, and validated the power tracking capability of said SPD. Procedural details of the experiments, data, calculations and associated graphs, are presented and discussed. (author)

  15. Feedback from dismantling operations (level 2) on EDF's first generation reactors

    International Nuclear Information System (INIS)

    West, J P.; Dionisio-Gomes, A.; Kus, J P.; Mervaux, P.; Bernet, P.; Dalmas, R.

    2003-01-01

    EDF's policy as regards the dismantling of the reactors that have ceased commercial operation, namely the eight power plants of the first generation and the Creys-Malville power plant, is explained. Generally speaking, prior to the year 2001, EDF had opted for the de-construction of these power plants to comply with a 'long wait' scenario, which consisted of waiting for a period of 5 to 10 years to achieve IAEA level 2 (partial release of the site), then postponing the total de-construction of the facilities for 25 to 50 years. Today, EDF has decided to undertake the total de-construction of these reactors, which have ceased commercial operation, over a period of 25 years. The purpose of this document is to present: - The reactors concerned, their background and their 'regulatory' situation, - The main operations performed and/or currently in progress, - The main elements of feedback from such operations, shedding light on the approach adopted in 2001. The installations concerned by the de-construction programme are as follows: - The 8 power plants of the first generation, which were built during the fifties and sixties and ceased commercial operation between 1973 and 1994, namely: Brennilis (industrial prototype using heavy water technology, jointly operated by EDF and CEA), the 6 power units of the NUGG type (natural uranium gas graphite) at Chinon, Saint-Laurent des Eaux and Bugey and the PWR reactor at Chooz A, - The storage silos at Saint-Laurent, where the sleeves for the fuel assemblies of reactors SLA1 and SLA2 are stored, corresponding to approximately 2000 tonnes of graphite, - The Creys-Malville reactor, FBR (fast breeder reactor) shut down in accordance with a government decision, which is currently undergoing decommissioning. At the current stage, our feedback from the dismantling operations carried out on nuclear facilities is based on (i) the work carried out or in progress that will make it possible to achieve the equivalent of IAEA level 2 in the

  16. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. Recovery of MARICO-2 sample part

    International Nuclear Information System (INIS)

    Ashida, Takashi; Ito, Hideaki

    2015-01-01

    At Joyo reactor MK-III core in May 2007, due to the design deficiencies of the disconnect mechanism of the holding part and the sample part of the experimental apparatus with instrumentation lines (MARICO-2), a disconnect failure incident occurred in the sample part after irradiation test. The deformation of the sample part due to this failure incurred its interference with the lower surface of reactor core upper structure and the holddown axis body. By this, the operating range of the rotary plug was restricted, leading to the partial inhibition of the fuel exchange function that precluded the access to 1/4 of the assemblies of the reactor core. In face of restoration work, the preparation for restoration such the exchange of upper core structure, and the recovery of MARICO-2 sample part are under way. The following items are introduced here: (1) summary of restoration work and overall process of restoration work, (2) recovery operation of MARICO-2 sample part, (3) exchange of the upper core structure that was conducted this year, and (4) results of recovery of MARIKO-2 sample part. (A.O.)

  17. A thermal hydraulic analysis in PWR reactors with UO2 or (U-Th)O2 fuel rods employing a simplified code

    International Nuclear Information System (INIS)

    Santos, Thiago A. dos; Maiorino, José R.; Stefanni, Giovanni L. de

    2017-01-01

    In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and safety parameters are respected. Considering this issue, this research aims to evaluate the APTh-100 reactor thermal limits. This PWR is a project developed in Universidade Federal do ABC (UFABC) using fuel composed of Uranium and Thorium oxide mixed (U,Th)O 2 . For this purpose, a simplified, although conservative, code was developed in a MATLAB environment named STC-MOX-Th 'Simplified Thermal-hydraulics Code-Mixed Oxide Thorium'. This code provides axial and radial temperature distribution, as well as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities, such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O 2 .The software uses basic laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation, considering the thermal conductive coefficient as a temperature function. In order to solve this equation, the finite elements method was used. Furthermore, the proportion of 36% of UO 2 was used to evaluate the temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middle and ending. The program has proven to be efficient in every condition and the results evidenced that the APTh-1000 reactor, in an initial analysis, has its thermal limits within the recommended security parameters. (author)

  18. 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

  19. 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.)

  20. Criticality analysis for mixed thorium-uranium fuel in the Angra-2 PWR reactor using KENO-VI

    Energy Technology Data Exchange (ETDEWEB)

    Wichrowski, Caio C.; Gonçalves, Isadora C.; Oliveira, Claudio L.; Vellozo, Sergio O.; Baptista, Camila O., E-mail: wichrowski@ime.eb.br, E-mail: isadora.goncalves@ime.eb.br, E-mail: d7luiz@yahoo.com.br, E-mail: vellozo@ime.eb.br, E-mail: camila.oliv.baptista@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Seção de Engenharia Nuclear

    2017-07-01

    The increasing energy demand associated to the current sustainability challenges have given the thorium nuclear fuel cycle renewed interest in the scientific community. Studies have focused on energy production in different reactor designs through the fission of uranium 233, the product of thorium fertilization by neutrons. In order to make it possible for near future applications a strategy based on the adaptation of current nuclear reactors for the use of thorium fuels is being considered. In this work, bearing in mind these limitations, a code was used to evaluate the effect on criticality (k{sub inf}) of the mixing of thorium and uranium in different proportions in the fuel of a PWR, the German designed Angra-2 Brazilian reactor in order to scrutinise its behaviour and determine the feasibility of an adapted ThO{sub 2}-UO{sub 2} mixed fuel cycle using current PWR technology. The analysis is performed using the KENO-VI module in the SCALE 6.1 nuclear safety analysis simulation code and the information is taken from the Angra-2 FSAR (Final Security Analysis Report). (author)

  1. Application of CO{sub 2} selective membrane reactors in pre-combustion decarbonisation systems for power production

    Energy Technology Data Exchange (ETDEWEB)

    Steven C.A. Kluiters; Virginie C. Feuillade; Jan Wilco Dijkstra; Daniel Jansen; Wim G. Haije [Energy research Centre of the Netherlands (ECN), Petten (Netherlands)

    2006-07-01

    For pre-combustion decarbonisation of fuels for large-scale power production or H{sub 2} generation both CO{sub 2} and H{sub 2} selective membranes are viable candidates for use in steam reforming and water gas shift membrane reactors. It will be shown that the choice between either option is not a matter of taste, but dictated by the fuel used and, to a lesser extent, the total system layout. Hydrotalcites, clay-like materials, are shown to be promising candidates as membrane material for low temperature, below 400{sup o}C, membrane shift reactors. 7 refs., 6 figs., 1 tab.

  2. Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1990-01-01

    Argonne National Laboratory's Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to >15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel

  3. CO2 Reduction Assembly Prototype Using Microlith-Based Sabatier Reactor for Ground Demonstration

    Science.gov (United States)

    Junaedi, Christian; Hawley, Kyle; Walsh, Dennis; Roychoudhury, Subir; Abney, Morgan B.; Perry, Jay L.

    2014-01-01

    The utilization of CO2 to produce life support consumables, such as O2 and H2O, via the Sabatier reaction is an important aspect of NASA's cabin Atmosphere Revitalization System (ARS) and In-Situ Resource Utilization (ISRU) architectures for both low-earth orbit and long-term manned space missions. Carbon dioxide can be reacted with H2, obtained from the electrolysis of water, via Sabatier reaction to produce methane and H2O. Methane can be stored and utilized as propellant while H2O can be either stored or electrolyzed to produce oxygen and regain the hydrogen atoms. Depending on the application, O2 can be used to replenish the atmosphere in human-crewed missions or as an oxidant for robotic and return missions. Precision Combustion, Inc. (PCI), with support from NASA, has previously developed an efficient and compact Sabatier reactor based on its Microlith® catalytic technology and demonstrated the capability to achieve high CO2 conversion and CH4 selectivity (i.e., =90% of the thermodynamic equilibrium values) at high space velocities and low operating temperatures. This was made possible through the use of high-heat-transfer and high-surface-area Microlith catalytic substrates. Using this Sabatier reactor, PCI designed, developed, and demonstrated a stand-alone CO2 Reduction Assembly (CRA) test system for ground demonstration and performance validation. The Sabatier reactor was integrated with the necessary balance-of-plant components and controls system, allowing an automated, single "push-button" start-up and shutdown. Additionally, the versatility of the test system prototype was demonstrated by operating it under H2-rich (H2/CO2 of >4), stoichiometric (ratio of 4), and CO2-rich conditions (ratio of <4) without affecting its performance and meeting the equilibrium-predicted water recovery rates. In this paper, the development of the CRA test system for ground demonstration will be discussed. Additionally, the performance results from testing the system at

  4. Sophistication of burnup analysis system for fast reactor (2)

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hirai, Yasushi; Tatsumi, Masahiro

    2010-10-01

    Improvement on prediction accuracy for neutronics characteristics of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified cross-section set as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores improvement of not only static characteristics but also burnup characteristics is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup characteristics using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for power reactor analysis systems. In the present study, by extending the prototype system, features for handling of control rods and energy collapse of group constants have been designed and implemented. Computational results from the present analysis system are stored into restart files which can be accessible by

  5. Modernization of turbine control system and reactor control system in Almaraz 1 and 2

    International Nuclear Information System (INIS)

    Pulido, C.; Diez, J.; Carrasco, J. A.; Lopez, L.

    2005-01-01

    The replacement of the Turbine Control System and Reactor Control System are part of the Almaraz modernization program for the Instrumentation and Control. For these upgrades Almaraz has selected the Ovation Platform that provides open architecture and easy expansion to other systems, these platforms is highly used in many nuclear and thermal plants around the world. One of the main objective for this project were to minimize the impact on the installation and operation of the plant, for that reason the project is implemented in two phases, Turbine Control upgrade and Reactor Control upgrade. Another important objective was to increase the reliability of the control system making them fully fault tolerant to single failures. The turbine Control System has been installed in Units 1 and 2 while the Reactor Control System will be installed in 2006 and 2007 outages. (Author)

  6. Outline of construction planning on No. 2 Reactor of the Shika Nuclear Power Plant

    International Nuclear Information System (INIS)

    Nakagawa, Tetsuro; Kadoki, Shuichi; Kubo, Tetsuji

    1999-01-01

    The Hokuriku Electric Co., Ltd. carries out the expansion of the Shika Nuclear Power Plant No.2 (ABWR) to start its in March 2006. It is situated in north neighboring side of No. 1 reactor under operation at present, and its main buildings are planned to position a reactor building at mountain side and a turbine building at sea side as well as those in the No. 1 reactor. And, cooling water for steam condenser was taken in from an intake opening built at north side of the lifting space situated at the front of the power plant, and discharged into seawater from a flashing opening positioned about 600 m offing. Here were described on outline of main civil engineering such as base excavation engineering, concrete caisson production, oceanic establishment engineering, and facility for steam condenser, and characteristics of the engineering. (G.K.)

  7. 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

  8. ENSI's view on technical safety for the long term operation of reactors 1 and 2 in the Beznau nuclear power plant

    International Nuclear Information System (INIS)

    2010-11-01

    The reactors 1 and 2 of the Beznau nuclear power plant (KKB) are operated since about 40 years. For an operation beyond the design period of 40 years the Swiss Federal Nuclear Safety Inspectorate (ENSI) demands the evidence to be brought that the design limits of the safety relevant components will not be reached during the extended operation period. In 2008 the license holder of KKB delivered the requested documentation on material ageing on the basis of deterministic as well as probabilistic safety analyses and concluded that both reactors can be safely operated beyond 40 years. Thanks to continuous additional outfits, both reactors are in good condition from the point of view of technical safety. With a view to the extension of operation beyond 40 years, KKB already applied the necessary measures regarding technics, finances and personnel in order to keep the present technical level. Since 1991 KKB has analysed and checked components that are difficult to replace. From the evidence presented, ENSI concluded that both reactors are able to be operated up to 60 years long, however with two restrictions for reactor 1 because there the material used for the reactor pressure vessel (RPV) suffered more neutron brittleness than in reactor 2. In addition, reactor 1 is much more affected by ageing phenomena than reactor 2, but, according to neutron fluence calculations, the limiting criteria will not be reached even after 60 years of operation. Some corrosion damages were noted at the lower part of the RPV due to water containing boron acid; they are more pronounced in reactor 1 than in reactor 2. Even though the calculations done by KKB are very conservative, they show that also in the long term the operation limiting criteria about the mechanical resistance of the RPV are never reached. ENSI concludes that the safety design of both KKB reactors ensures safe control of the design basis accidents. Both reactors were continuously fitted with new equipment. With the planed

  9. Ceramics as nuclear reactor fuels

    International Nuclear Information System (INIS)

    Reeve, K.D.

    1975-01-01

    Ceramics are widely accepted as nuclear reactor fuel materials, for both metal clad ceramic and all-ceramic fuel designs. Metal clad UO 2 is used commercially in large tonnages in five different power reactor designs. UO 2 pellets are made by familiar ceramic techniques but in a reactor they undergo complex thermal and chemical changes which must be thoroughly understood. Metal clad uranium-plutonium dioxide is used in present day fast breeder reactors, but may eventually be replaced by uranium-plutonium carbide or nitride. All-ceramic fuels, which are necessary for reactors operating above about 750 0 C, must incorporate one or more fission product retentive ceramic coatings. BeO-coated BeO matrix dispersion fuels and silicate glaze coated UO 2 -SiO 2 have been studied for specialised applications, but the only commercial high temperature fuel is based on graphite in which small fuel particles, each coated with vapour deposited carbon and silicon carbide, are dispersed. Ceramists have much to contribute to many aspects of fuel science and technology. (author)

  10. Novel, Regenerable Microlith Catalytic Reactor for CO2 Reduction via Bosch Process, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Precision Combustion, Inc. (PCI) proposes to develop an extremely compact, lightweight and regenerable MicrolithREG catalytic CO2 reduction reactor, capable of...

  11. Advanced reactor development

    International Nuclear Information System (INIS)

    Till, C.E.

    1989-01-01

    Consideration is given to what the aims of advanced reactor development have to be, if a new generation of nuclear power is really to play an important role in man's energy generation activities in a fragile environment. The background given briefly covers present atmospheric evidence, the current situation in nuclear power, how reactors work and what can go wrong with them, and the present magnitudes of world energy generation. The central part of the paper describes what is currently being done in advanced reactor development and what can be expected from various systems and various elements of it. A vigorous case is made that three elements must be present in any advanced reactor development: (1) breeding; (2) passive safety; and (3) shorter-live nuclear waste. All three are possible. In the right advanced reactor systems the ways of achieving them are known. But R and D is necessary. That is the central argument made in the paper. Not advanced reactor prototype construction at this point, but R and D itself. (author)

  12. Reactor power control device

    International Nuclear Information System (INIS)

    Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

    1998-01-01

    The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

  13. Final report on in-reactor creep-fatigue deformation behaviour of a CuCrZr alloy: COFAT 2

    International Nuclear Information System (INIS)

    Singh, B.N.; Johansen, B.S.; Taehtinen, S.; Moilanen, P.; Saarela, S.; Jacquet, P.; Dekeyser, J.; Stubbins, J.F.

    2008-01-01

    The main objective of the present work was to determine experimentally the mechanical response and resulting microstructural changes in CuCrZr (HT1) alloy exposed concurrently to flux of neutrons and creep-fatigue cyclic loading directly in a fission reactor. Using specially designed test facilities for this purpose, in-reactor creep-fatigue tests have been performed at strain amplitudes of 0.25 and 0.35 % with a holdtime of 10s in the BR-2 reactor at Mol (Belgium). These tests were performed at the ambient temperatures of 326K and 323K. For comparison purposes corresponding out-of-reactor creep-fatigue tests were also carried out. In the following we first describe the details of the creep-fatigue experiments. We then present the main results on the mechanical response of the material in the form of hysteresis loops and the maximum stress amplitude as a function of the number of creep-fatigue cycles during the out-of-reactor and the in-reactor tests carried out at different strain amplitudes. Finally, the dependence of the number of cycles to failure (i.e. creep-fatigue lifetime) on the strain amplitudes is shown. The details of microstructure of the specimens tested out-of-reactor as well as in the reactor were investigated using transmission electron microscopy. The main results on the mechanical response as well as changes in the microstructure are briefly discussed. The main conclusion emerging from the present work is that the lifetime of the in-reactor tested specimens is by a factor of about two longer than in the case of corresponding out-of-reactor tests. (au)

  14. Propagation of negative electrical discharges through 2-dimensional packed bed reactors

    International Nuclear Information System (INIS)

    Kruszelnicki, Juliusz; Engeling, Kenneth W; Foster, John E; Xiong, Zhongmin; Kushner, Mark J

    2017-01-01

    Plasma-based pollutant remediation and value-added gas production have recently gained increased attention as possible alternatives to the currently-deployed chemical reactor systems. Electrical discharges in packed bed reactors (PBRs) are of interest, due to their ability to synergistically combine catalytic and plasma chemical processes. In principle, these systems could be tuned to produce specific products, based on their application by combinations of power formats, materials, geometries and working gases. Negative voltage, atmospheric-pressure plasma discharges sustained in humid air in a PBR-like geometry were experimentally characterized using ICCD imaging and simulated in 2-dimensions (2D) to provide insights into possible routes to this tunability. Surface ionization waves (SIWs) and positive restrikes through the lattice of dielectric rods were shown to be the principal means of producing reactive species. The number and intensity of SIWs and restrikes are sensitive functions of the alignment of the lattice of dielectric beads (or rods in 2D) with respect to the applied electric field. Decreased spacing between the dielectric elements leads to an increased electric field enhancement in the gas, and therefore locally higher plasma densities, but does not necessarily impact the types of discharges that occur through the lattice. (paper)

  15. Propagation of negative electrical discharges through 2-dimensional packed bed reactors

    Science.gov (United States)

    Kruszelnicki, Juliusz; Engeling, Kenneth W.; Foster, John E.; Xiong, Zhongmin; Kushner, Mark J.

    2017-01-01

    Plasma-based pollutant remediation and value-added gas production have recently gained increased attention as possible alternatives to the currently-deployed chemical reactor systems. Electrical discharges in packed bed reactors (PBRs) are of interest, due to their ability to synergistically combine catalytic and plasma chemical processes. In principle, these systems could be tuned to produce specific products, based on their application by combinations of power formats, materials, geometries and working gases. Negative voltage, atmospheric-pressure plasma discharges sustained in humid air in a PBR-like geometry were experimentally characterized using ICCD imaging and simulated in 2-dimensions (2D) to provide insights into possible routes to this tunability. Surface ionization waves (SIWs) and positive restrikes through the lattice of dielectric rods were shown to be the principal means of producing reactive species. The number and intensity of SIWs and restrikes are sensitive functions of the alignment of the lattice of dielectric beads (or rods in 2D) with respect to the applied electric field. Decreased spacing between the dielectric elements leads to an increased electric field enhancement in the gas, and therefore locally higher plasma densities, but does not necessarily impact the types of discharges that occur through the lattice.

  16. Turning points in reactor design

    International Nuclear Information System (INIS)

    Beckjord, E.S.

    1995-01-01

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems

  17. Turning points in reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Beckjord, E.S.

    1995-09-01

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

  18. Mimic of OSU research reactor

    International Nuclear Information System (INIS)

    Lu, Hong; Miller, D.W.

    1991-01-01

    The Ohio State University research reactor (OSURR) is undergoing improvements in its research and educational capabilities. A computer-based digital data acquisition system, including a reactor system mimic, will be installed as part of these improvements. The system will monitor the reactor system parameters available to the reactor operator either in digital parameters available to the reactor operator either in digital or analog form. The system includes two computers. All the signals are sent to computer 1, which processes the data and sends the data through a serial port to computer 2 with a video graphics array VGA monitor, which is utilized to display the mimic system of the reactor

  19. Evaluation on activation activity of reactor in JRR-2 applied 3 dimensional model to neutron flux calculation

    International Nuclear Information System (INIS)

    Kishimoto, Katsumi; Arigane, Kenji

    2005-03-01

    Revaluation to activation activity of reactor evaluated at the notification of dismantling submitted in 1997 was carried out in JRR-2 where decommissioning was advanced now. In the revaluation, estimation accuracy on neutron streaming at various horizontal experimental tubes was improved by applying 3 dimensional model to neutron transport calculation that had been carried out by 2 dimensional model, and calculating with TORT. As the result, excessive overestimations on horizontal experimental tubes and biological shield that had greatly contributed to total activation activity in evaluation at the notification of dismantling was revised, sum of their activation activities in the revaluation decreased to 1/18 (case after 1 year from the permanent shutdown of reactor) of evaluation at the notification of dismantling, and the structural materials that had large activation activity were changed. By the above, it was shown that introducing 3 dimensional model was effective in evaluation on activation activity of the research reactor that had a lot of various experimental tubes. Total activation activity of reactor by the revaluation depended on control rods, thermal shield plates and horizontal experimental tubes, and the value after 1 year from the permanent shutdown of reactor was 1.9x10 14 Bq. (author)

  20. In-pile modelling of nuclear fuel element for the MTR type reactors. Pt. 2

    Energy Technology Data Exchange (ETDEWEB)

    Farhadi, Kazem [AEOI, Tehran (Iran, Islamic Republic of). Radiations Application Research School

    2014-06-15

    In part two of the present paper, neutronic properties of the pool-type research reactor core are used to assess the similitude laws derived for out-of-pile modelling of the fuel element. The benchmark reactor used for this purpose is an IAEA 5 MW thermal pool-type research reactor currently in operation. The neutronic properties analysis are based on typical 2 200 m/sec and neutrons having 0.025 eV energy. The non-leakage capability of the system is estimated in terms of diffusion length. Also the slowing down power and the moderating ratio of the modelled methanol coolant are calculated in terms of lethargy of the diffusing medium. It is shown that the Iron which is substituted for Aluminium cladding is a relatively low absorber of neutrons but has a high neutron leakage. Methanol which replaced ordinary water as coolant is not a suitable coolant due to high neutrons absorbing substance. It is concluded that although Iron as a cladding material and methanol as a coolant meet the modelling out-of-pile criteria but are not satisfying neutronic properties. Therefore, use of them as a model clad and coolant are not suggested for research reactors. (orig.)

  1. Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1976-09-01

    Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO 2 ) and lithium silicate (Li 2 SiO 3 ) by the reaction: Li 6 + n → 4 He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100 0 C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T 2 ), while in laboratory extractions (300-1300 0 C), the tritium appeared primarily in the condensible form (HTO and T 2 O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H 2 O, CO 2 , CO, O 2 , H 2 , NO, SO 2 , SiF 4 and traces of hydrocarbons

  2. A KINETIC MODEL FOR H2O2/UV PROCESS IN A COMPLETELY MIXED BATCH REACTOR. (R825370C076)

    Science.gov (United States)

    A dynamic kinetic model for the advanced oxidation process (AOP) using hydrogen peroxide and ultraviolet irradiation (H2O2/UV) in a completely mixed batch reactor (CMBR) is developed. The model includes the known elementary chemical and photochemical reac...

  3. Installation modification of the reactor No.2 of Ikata nuclear power plant of Shikoku Electric Power Company, Inc

    International Nuclear Information System (INIS)

    1980-01-01

    The application was made on August 25, 1979, from the president of the Shikoku Electric Power Company, Inc., to the Minister of International Trade and Industry, relating to the installation modification of the reactor No. 2 in the Ikata nuclear power plant. The inquiry was submitted on September 28, 1979, from the Minister of International Trade and Industry to the Nuclear Safety Commission, after the safety evaluation in the Ministry of International Trade and Industry, and the investigation and deliberation were started on October 1, 1979, in the Nuclear Safety Commission. The content of the modification is to add the circuit actuated by the abnormal low pressure signal of the reactor to the actuating circuit of the emergency core cooling system (ECCS) and to increase the new fuel storage capacity from about 1/3 core to about 2/3 core. The additional signal circuit is composed of the logic circuit of ''2 out of 4'' and is multichannel design. The circuit is independent from the reactor control system and the conventional signal circuit of the concurrence of low pressure in the reactor and low level in the pressurizer. With the addition of the circuit of abnormal low pressure signal of the reactor, the countermeasures for preventing ECCS start by mistake are also added. These modifications give no influence to the functions of the reactor control system and reactor protection system. The function and the performance of ECCS were analyzed and evaluated accompanying these modifications assuming the loss of coolant accident. Concerning the new fuel storage capacity, the type of racks is modified from angle type to can type, and the subcriticality is kept even at the time of water flood. (Nakai, Y.)

  4. Reactor Structure Materials: Corrosion of Reactor Core Internals

    International Nuclear Information System (INIS)

    Van Dyck, S.

    2000-01-01

    The objectives of SCK-CEN's R and D programme on the corrosion of reactor core internals are: (1) to gain mechanistic insight into the Irradition Assisted Stress Corrosion Cracking (IASCC) phenomenon by studying the influence of separate parameters in well controlled experiments; (2) to develop and validate a predictive capability on IASCC by model description and (3) to define and validate countermeasures and monitoring techniques for application in reactors. Progress and achievements in 1999 are described

  5. Uranium redistribution under oxidizing conditions in Oklo natural reactor zone 2, Gabon

    International Nuclear Information System (INIS)

    Isobe, H.; Ohnuki, T.; Murakami, T.; Gauthier-Lafaye, F.

    1995-01-01

    This mineralogical study was completed to elucidate the relationships between uranium distribution and alteration products of the host rock of natural reactor zone clays just below the reactor core. Uraninite is preserved without any alteration in the reactor core. Uranium minerals are found to be present in the fractures in the reactor zone clays associated with iron-mineral veins, galena and Ti-bearing minerals. Uranium, for which the phases could not be identified, occurs in iron-mineral veins and the iron-mineral rim of pyrite grains in the reactor zone clays. Uranium is not associated with granular iron minerals occurring in the illite matrix of the reactor zone clays. The degree of crystallinity and uranium content of the three iron-bearing alteration products suggest that they formed under different conditions; the granular iron minerals, under alteration conditions where uranium was not mobilized while the iron-mineral veins and the iron-mineral rim of pyrite, under conditions in which uranium is mobilized after the formation of the granular iron minerals

  6. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 1. MARICO-2 subassembly retrieval work

    International Nuclear Information System (INIS)

    Naito, Hiroyuki; Ashida, Takashi; Ito, Hideaki

    2014-01-01

    At Joyo reactor MK-III core in May 2007, due to the design deficiencies of the disconnect mechanism of the holding part and the sample part of the experimental apparatus with instrumentation lines (MARICO-2), a disconnect failure incident occurred in the sample part after irradiation test. The deformation of the sample part due to this failure incurred its interference with the lower surface of reactor core upper structure and the holddown axis body. By this, the operating range of the rotary plug was restricted, leading to the partial inhibition of the fuel exchange function that precluded the access to 1/4 of the assemblies of the reactor core. In face of restoration work, the preparation for restoration such the exchange of upper core structure, and the recovery of MARICO-2 sample part are under way. This paper introduces the progress of restoration work and the future work plan, with a focus on the outline of overall restoration work, the method / problems / measures for MARICO-2 sample part recovery operations, and fabrication of sample part recovery device. (A.O.)

  7. Calculation of photon dose for Dalat research reactor in case of loss of reactor tank water

    International Nuclear Information System (INIS)

    Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2007-01-01

    Photon sources of actinides and fission products were estimated by ORIGEN2 code with the modified cross-section library for Dalat research reactor (DRR) using new cross-section generated by WIMS-ANL code. Photon sources of reactor tank water calculated from the experimental data. MCNP4C2 with available non-analog Monte Carlo model and ANSI/ANL-6.1.1-1977 flux-to-dose factors were used for dose estimation. The agreement between calculation results and those of measurements showed that the methods and models used to get photon sources and dose were acceptable. In case the reactor water totally leaks out from the reactor tank, the calculated dose is very high at the top of reactor tank while still low in control room. In the reactor hall, the operation staffs can access for emergency works but with time limits. (author)

  8. Countertransference when working with narcissistic personality disorder: An empirical investigation.

    Science.gov (United States)

    Tanzilli, Annalisa; Muzi, Laura; Ronningstam, Elsa; Lingiardi, Vittorio

    2017-06-01

    Narcissistic personality disorder (NPD) is one of the most challenging clinical syndromes to treat in psychotherapy, especially due to the difficulties of establishing a good enough therapist-patient relationship. Countertransference responses to NPD can be particularly intense, frustrating, and difficult to manage, as is often reported in the clinical literature though not clearly supported empirically. The aims of this study were to (a) investigate the relationship between patients' NPD and therapists' responses; (b) examine the associations between patient, clinician, therapy variables and clinicians' reactions during treatment of NPD patients; and (c) provide an empirically derived portrait of countertransference with NPD. A sample of psychiatrists and clinical psychologists (N = 67) completed the Therapist Response Questionnaire to identify patterns of countertransference, the Shedler-Westen Assessment Procedure-200, and the Global Assessment of Functioning Scale to assess the personality pathology and psychosocial functioning of a patient in their care. The results showed that NPD was positively associated with hostile/angry, criticized/devalued, helpless/inadequate, and disengaged countertransference and negatively associated with therapists' positive response, regardless of patients' personality and psychosocial functioning. NPD patients with stronger traits of cluster B personality pathology tended to elicit more negative and heterogeneous countertransference reactions than NPD patients without these features. The countertransference patterns with NPD patients were not strongly influenced by the variables of clinicians and therapy, with the exception of clinical experience. Overall, the portrait of therapists' reactions to NPD provided a clinically nuanced and empirically founded description strongly resembling theoretical-clinical accounts. The therapeutic implications of these findings were discussed. (PsycINFO Database Record (c) 2017 APA, all rights

  9. Review of mirror fusion reactor designs

    International Nuclear Information System (INIS)

    Bender, D.J.

    1977-01-01

    Three magnetic confinement concepts, based on the mirror principle, are described. These mirror concepts are summarized as follows: (1) fusion-fission hybrid reactor, (2) tandem mirror reactor, and (3) reversed field mirror reactor

  10. Visible-Light-Responsive Photocatalysis: Ag-Doped TiO2 Catalyst Development and Reactor Design Testing

    Science.gov (United States)

    Coutts, Janelle L.; Hintze, Paul E.; Meier, Anne; Shah, Malay G.; Devor, Robert W.; Surma, Jan M.; Maloney, Phillip R.; Bauer, Brint M.; Mazyck, David W.

    2016-01-01

    In recent years, the alteration of titanium dioxide to become visible-light-responsive (VLR) has been a major focus in the field of photocatalysis. Currently, bare titanium dioxide requires ultraviolet light for activation due to its band gap energy of 3.2 eV. Hg-vapor fluorescent light sources are used in photocatalytic oxidation (PCO) reactors to provide adequate levels of ultraviolet light for catalyst activation; these mercury-containing lamps, however, hinder the use of this PCO technology in a spaceflight environment due to concerns over crew Hg exposure. VLR-TiO2 would allow for use of ambient visible solar radiation or highly efficient visible wavelength LEDs, both of which would make PCO approaches more efficient, flexible, economical, and safe. Over the past three years, Kennedy Space Center has developed a VLR Ag-doped TiO2 catalyst with a band gap of 2.72 eV and promising photocatalytic activity. Catalyst immobilization techniques, including incorporation of the catalyst into a sorbent material, were examined. Extensive modeling of a reactor test bed mimicking air duct work with throughput similar to that seen on the International Space Station was completed to determine optimal reactor design. A bench-scale reactor with the novel catalyst and high-efficiency blue LEDs was challenged with several common volatile organic compounds (VOCs) found in ISS cabin air to evaluate the system's ability to perform high-throughput trace contaminant removal. The ultimate goal for this testing was to determine if the unit would be useful in pre-heat exchanger operations to lessen condensed VOCs in recovered water thus lowering the burden of VOC removal for water purification systems.

  11. On the anisotropy of in-reactor creep of Zr-2.5Nb tubes

    International Nuclear Information System (INIS)

    Causey, A.R.; Holt, R.A.

    1993-06-01

    Creep specimens made from cold-worked Zr-2.5Nb tubes, fabricated with two different microstructures and crystallographic textures, were irradiated in the Osiris reactor in France in a fast-neutron flux of about 1.8 x 10 18 n.m -2 .s -1 , E > MeV, at 553 and 585 K. The hoop stresses from internal Fluences, up to 4 x 10 25 n.m -2 , more than double those achieved an any other creep test on cold-worked Zr-2.5Nb in which both axial and transverse strain were measured. Creep rates were obtained from strain versus fluence plots, and creep compliances were obtained from plots of the strain rates against hoop stress for each material at each temperature. The ratio of creep rates at 583 K to those at 553 K was ∼ 1.36, a little higher than that extrapolated from stress relaxation results at temperatures between 523 and 568 K. The ratio of the biaxial creep compliance in the axial direction to that in the transverse directions is different for the two test materials: 0.0 to -0.1 for the fuel sheathing texture and 0.5 to 0.6 for the pressure tube texture. The results were analysed using a self-consistent model developed to account for the contributions to the creep anisotropy of the three microstructure parameters involved and to account for the grain interaction effects. The model, which was normalized to test reactor and power reactor creep data for cold-worked Zr-2.5Nb tubes, predicted the ratio of the creep compliancies to be -0.26 and 0.63, respectively. Thus the creep anisotropy of Zr-2.5Nb tubes with pressure-tube-like crystallographic texture can be adequately predicted. (author). 18 refs., 4 tabs., 13 figs

  12. COSTANZA, 1-D 2 Group Space-Dependent Reactor Dynamics of Spatial Reactor with 1 Group Delayed Neutrons

    International Nuclear Information System (INIS)

    Agazzi, A.; Gavazzi, C.; Vincenti, E.; Monterosso, R.

    1964-01-01

    1 - Nature of physical problem solved: The programme studies the spatial dynamics of reactor TESI, in the two group and one space dimension approximation. Only one group of delayed neutrons is considered. The programme simulates the vertical movement of the control rods according to any given movement law. The programme calculates the evolution of the fluxes and temperature and precursor concentration in space and time during the power excursion. 2 - Restrictions on the complexity of the problem: The maximum number of lattice points is 100

  13. Lack of empathy in patients with narcissistic personality disorder.

    Science.gov (United States)

    Ritter, Kathrin; Dziobek, Isabel; Preissler, Sandra; Rüter, Anke; Vater, Aline; Fydrich, Thomas; Lammers, Claas-Hinrich; Heekeren, Hauke R; Roepke, Stefan

    2011-05-15

    The study's objective was to empirically assess cognitive and emotional empathy in patients with narcissistic personality disorder (NPD). To date, "lack of empathy" is a core feature of NPD solely based on clinical observation. The study's method was that forty-seven patients with NPD, 53 healthy controls, and 27 clinical controls with borderline personality disorder (BPD) were included in the study. Emotional and cognitive empathy were assessed with traditional questionnaire measures, the newly developed Multifaceted Empathy Test (MET), and the Movie for the Assessment of Social Cognition (MASC). The study's results were that individuals with NPD displayed significant impairments in emotional empathy on the MET. Furthermore, relative to BPD patients and healthy controls, NPD patients did not show deficits in cognitive empathy on the MET or MASC. Crucially, this empathic profile of NPD is not captured by the Structured Clinical Interview for DSM-IV for Axis II Disorders (SCID-II). The study's conclusions were that while NPD involves deficits in emotional empathy, cognitive empathy seems grossly unaffected. Copyright © 2010 Elsevier Ltd. All rights reserved.

  14. Present status of reactor physics in the United States and Japan-IV. 2. Micro-Reactor Physics of MOX-Fueled Core

    International Nuclear Information System (INIS)

    Takeda, Toshikazu

    2001-01-01

    Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design. We used the subgroup method to treat the space dependence of the self-shielding effect of heavy nuclides, and we used the characteristics method to treat the angular dependence of neutron flux in a fuel pellet. Figure 1 compares the power distributions in MOX and UO 2 fuel cells at the beginning of burnup. The power is calculated with and without considering the space dependence of the self-shielding effect of the cross sections. For the MOX cell, the power distribution has a peak at the cell edge because of large Pu absorption especially when considering the spatial self-shielding effect. When a MOX rod is adjacent to UO 2 fuel rods, the flux distribution has an azimuthal dependence in addition to the radial dependence within a rod. For example, consider a 2x2 fuel assembly composed of three UO 2 rods and one MOX rod, with the mirror reflection boundary condition. A burnup calculation was done with the condition; the radius of the MOX pellet is divided into two regions, and the azimuthal angle is divided into eight. The number density of 239 Pu at 44 000 MWd/t for the MOX rod shows azimuthal dependence by 20%. The maximum burnup occurs in the direction of the UO 2 rods. This is

  15. Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

    International Nuclear Information System (INIS)

    Čerba, Štefan; Vrban, Branislav; Lüley, Jakub; Dařílek, Petr; Zajac, Radoslav; Nečas, Vladimír; Haščik, Ján

    2014-01-01

    Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice

  16. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guigon, B. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, Reacteur Jules Horowitz, 13 - Saint-Paul-lez-Durance (France); Vacelet, H. [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, CERCA, Etablissement de Romans, 26 (France); Dornbusch, D. [Technicatome, Service d' Architecture Generale, 13 - Aix-en-Provence (France)

    2003-07-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs: from activation analysis to power reactor fuel qualification. In this paper will be presented the main characteristics of the Jules Horowitz Reactor: its total power, neutron flux, fuel element... Safety criteria will be explained. Finally merits and disadvantages of UMo compared to the standard U{sub 3}Si{sub 2} fuel will be discussed. (authors)

  17. Reactor Materials Research

    International Nuclear Information System (INIS)

    Van Walle, E.

    2001-01-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  18. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  19. Final report on in-reactor creep-fatigue deformation behaviour of a CuCrZr alloy: COFAT 2

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Johansen, Bjørn Sejr; Tähtinen, S.

    facilities for this purpose, in-reactor creep-fatigue tests have been performed at strain amplitudes of 0.25 and 0.35 % with a holdtime of 10s in the BR-2 reactor at Mol (Belgium). These tests were performed at the ambient temperatures of 326K and 323K. For comparison purposes corresponding out...

  20. Investigation of tritium and 233U breeding in a fission-fusion hybrid reactor fuelling with ThO2

    International Nuclear Information System (INIS)

    Yildiz, K.; Sahin, S.; Sahin, H. M.; Acir, A.; Yalcin, S.; Altinok, T.; Bayrak, M.; Alkan, M.; Durukan, O.

    2007-01-01

    In the world, thorium reserves are three times more than natural Uranium reserves. It is planned in the near future that nuclear reactors will use thorium as a fuel. Thorium is not a fissile isotope because it doesn't make fission with thermal neutrons so it could be converted to 2 33U isotope which has very high quality fission cross-section with thermal neutrons. 2 33U isotope can be used in present LWRs as an enrichment fuel. In the fusion reactors, tritium is the most important fossil fuel. Because tritium is not natural isotope, it has to be produced in the reactor. The purpose of this work is to investigate the tritium and 2 33U breeding in a fission-fusion hybrid reactor fuelling with ThO 2 for Δt=10 days during a reactor operation period in five years. The neutronic analysis is performed on an experimental hybrid blanket geometry. In the center of the hybrid blanket, there is a line neutron source in a cylindrical cavity, which simulates the fusion plasma chamber where high energy neutrons (14.1 MeV) are produced. The conventional fusion reaction delivers the external neutron source for blankets following, 2 D + 3 T →? 4 He (3.5 MeV) + n (14.1 MeV). (1) The fuel zone made up of natural-ThO 2 fuel and it is cooled with different coolants. In this work, five different moderator materials, which are Li 2 BeF 4 , LiF-NaF-BeF 2 , Li 2 0Sn 8 0, natural Lithium and Li 1 7Pb 8 3, are used as coolants. The radial reflector, called tritium breeding zones, is made of different Lithium compounds and graphite in sandwich structure. In the work, eight different Lithium compounds were used as tritium breeders in the tritium breeding zones, which are Li 3 N, Li 2 O, Li 2 O 2 , Li 2 TiO 3 , Li 4 SiO 3 , Li 2 ZrO 3 , LiBr and LiH. Neutron transport calculations are conducted in spherical geometry with the help of SCALE4.4A SYSTEM by solving the Boltzmann transport equation with code CSAS and XSDRNPM, under consideration of unresolved and resolved resonances, in S 8 -P 3